ML20138P861
| ML20138P861 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Sequoyah  |
| Issue date: | 02/12/1997 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML082401529 | List: |
| References | |
| NUDOCS 9703050425 | |
| Download: ML20138P861 (73) | |
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ENCLOSURE 2 TENNEG8EE VALLEY AUTHORITY SEQUOYAB NUCLEAR PLANT (8QN)
UNITS 1 AND 2 I
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RESPON8E TO REQUEST FOR INFORNATION REGARDING ADEQUACY AND AVAILABILITY OF DESIGN BASE 8 INFORMATION i
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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 RESPON8E TO REQUEST FOR INFORMATION REGARDING ADEQUACY 1
AND AVAILABILITY OF DESIGN BASES INFORMATION TABLE OF CONTENTS Page I. EXECUTIVE
SUMMARY
1 II.
SUMMARY
OF SQN DESIGN BASIS VER'tFICATION EFFORTS...........................................
3 A.
Background...................................
3 B.
Design Baseline Verifica' ion Program (DBVP).......................................
3 C.
DBVP Implementation..........................
5 D.
Determination of Design Basis Requirements... 7 E.
Independent Reviews.........................
10 F.
Restart Test Program........................
11 G.
Summary....................
................ 12 III. Specific NRC Requested Information...............
13 A.
Request (a).................................
13 1.
Design and Configuration Control........ 13 a.
Design Change Control Process.......
14 b.
Fuel / Core Component Change Control............................. 16 c.
Temporary Alteration Control........
17 d.
Design Process Performance Monitoring..........................
18 2.
Activities that Maintain Design Configuration...........................
18 a.
System Line-up Controls.............
18 i
TABLE OF CONTENTS (Continued)
Page b.
Overall Control of the operation of Plant Equipment..................
1, c.
Control of Equipment During Maintenance and Modifications.......
19 d.
Post-modification / Post-maintenance Testing................. 20 e.
Control of Replacement Parts During Maintenance.........................
21 f.
Independent /Second-check / Concurrent Verification.......................*.
21 g.
Access to Design and Licensing Basis Information.........................
21 3.
10 CFR 50.59 Safety Assessments / Safety Evaluations............................. 22
- 4. Updated Final Safety Analysis Report (10 CFR 50.71(e)]........................
23
- 5. Implementation of Appendix B to 10 CFR Part 50...........................
25 6.
Commitment Management................... 26 B.
Request (b)
................................28 1.
Background.............................. 28 2.
Procedural Controls..................... 29 a.
Design Change Process...............
29 b.
Procedure Generation / Revision Process.............................
29 c.
Vendor Manual Control Process.......
30 3.
Procedure Verifications................. 30 a.
Procedure Upgrade Program...........
30 b.
Operating Experience Program........
33 ii
4 TABLE OF CONTENTS (Continued)
Page c.
Generic Regulatory Issues...........
34
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d.
Emergency Operating Procedures...... 35 4.
Independent Assessments................. 35 g
a.
Quality Assurance Assessments.......
36 b.
NRC Inspections.....................
36 5.
Summary................................. 37
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C.
Request (c).................................
38 1.
Configuration and Performance Controls................................ 38 j
2.
Verification Programs................... 38 a.
Reconciliation of Design Basis to Plant............................
39 b.
Restart Test Program................
39 3.
SSC Testing............................. 40 a.
Routine Surveillance Testing........
40 b.
ASME Code Required Performance Monitoring..........................
40 (1) Inservice Inspection............ 40 (2) System Pressure Tests........... 41 j
i (3) Inservice Testing............... 41 c.
Containment Testing.................
41 4.
Continuing Review Efforts............... 42 a.
Maintenance Rule Requirements.......
42 b.
System Health Reports...............
42 c.
Operating Experience Program........
42 d.
Generic Regulatory Issues...........
43 iii
TABLE OF CONTENTS (Continued)
Page 5.
Independent Assessments................. 44 a.
TVA Assessments.....................
44 u
b.
NRC Assessments.....................
45 i
l 6.
Summary................................. 45 i
i D.
Request (d)...................<.............
46 1.
TVA Corrective Action Program........... 47 a.
WR/WO Process.......................
47 4
b.
PER Process.........................
48 c.
Self-Assessments....................
52 d.
Operating Experience Program........ 53 e.
Event Reporting Process.............
54 l
f.
Informal Reporting..................
55 4
E.
Request (e)
................................56 i
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- 1. Summary for TVA Confidence in Design Basis................................... 56 I
2.
Performance / Implementation Issues....... 57 3.
Measurements of Effectiveness........... 58 i
a.
Design Control Process..............
58 b.
Corrective Action Program...........
59 c.
Design Baseline Verification
)
Program.............................
61 d.
Procedure Upgrade Program...........
62 l
e.
Site Management.....................
62
- 4. Updated Final Safety Analysis Report.... 63 l
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TABLE OF CONTENTS (Continued) i Page
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F.
Additional Efforts..........................
66 i
1.
FSAR Verification....................... 66 2.
Confirmatory Vertical Slice Review Assessments.............................
66 3.
Licensing Basis.........................
66 4.
Material Condition......................
67 5.
Corrective Action Program............... 67 i
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i ENCLOSURE 2 i
. TENNESSEE VALLEY AUTHORITY I
SEQUOYAE NUCLEAR PLANT (SQN)
UNITS 1 AND 2 i
RESPONSE TO REQUEST FOR INFORMATION REGARDING ADEQUACY AND AVAILABILITY OF DESIGN BASES INFORNATION
[
I.
EXECUTIVE SUNNARY This enclosure provides TVA's response to the i
October _9, 1996, 10 CFR 50.54(f) request for information regarding the adequacy, availability, and control of design i'
basis information for SQN.
As described in this enclosure, TVA has appropriate documentation defining the design basis for SQN.
TVA also has processes and controls designed to ensure that the design. basis, and changes made to the design basis, are appropriately evaluated and reflected in i
procedures as well as in system, structure and component i
(SSC) configuration and performance.
TVA has performed i
several programs which verified that the design basis has i
been translated into procedures and SSC configuration and performance, and we continue to perform reviews and assessments that evaluate this translation.
TVA's efforts in this area are not static, but rather are continually being improved and enhanced in light of ongoing industry and regulatory developments and TVA's own internal assessments.
When problems or the need for enhancements are identified, they are addressed through TVA's corrective action program.
r All of these programs, processes, controls, and reviews, as l
well as TVA's self-critical approach in this area, provide TVA with assurance that SQN is operated safely and v
l consistent with SQN design basis requirements.
i It was approximately ten years ago that TVA identified a failure at SQN to consistently maintain a documented design i
basis and to control the plant's configuration in accordance with that basis.
To address this concern, TVA implemented several improvement programs, including a Design Basis l
Verification Program (DBVP) for Units 1 and 2.
The scope of l
the DBVP included developing design documentation for those l
systems, or' portions thereof, that perform safety-related functions.
This. included the safety functions necessary to mitigate postulated design basis accidents, abnormal operational transients, or special events which are discussed in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR).
The DBVP also included plant walkdowns and l
drawing verfications that assessed plant configuration.
Once design requirements were developed through this program, they were incorporated into design documents, i.e.,
j design criteria, calculations, and drawings.
These p
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documents are reviewed when design changes are made and are updated as part of the configuration control process.
i TVA also implemented several programs to upgrade operations, maintenance, and testing procedures.
In addition, TVA implemented testing programs that validated the performance of SSCs.
These programs utilized the results of the DBVP to ensure design basis requirements were incorporated into the procedures.
Since the implementation of those programs, the integrity of TVA's design basis, as well as the translation of that basis into procedures, and SSC configuration and performance have been examined and tested through several mechanisms.
These mechanisms include TVA review efforts that target specific industry and regulatory issues, TVA i
l self-assessments, and NRC inspections.
TVA recognizes the importance of maintaining plant configuration consistent with the design basis and the need to control changes to the design basis to ensure that design basis assumptions remain valid.
TVA also recognizes the importance of maintaining an accurate UFSAR.
TVA continues l
to be self-critical in these areas and has identified the need for improvement in UFSAR accuracy.
Accordingly, TVA has initiated actions to review the UFSAR and the processes used to control the TT/SAR.
Over the past few yaars, TVA's aEsessments of design-related activities identifjed the following performance issues for which TVA has taken, or is in the process of taking, corrective actions:
plant activities that affected plant design, but that did not undergo the evaluation and review l
dictated by TVA's design control process; full close out of older design changes; timeliness of corrective actions; and proper dispositioning of abandoned equipment.
These performance issues have not resulted in the identification of conditions that would render safety-related systems inoperable nor have they undermined TVA's overall confidence in its design control process.
Details of the processes, programs and reviews of design basis information and its translation into procedures, and SSC configuration and performance are provided below.
Also included are TVA's plans for future activities in this area.
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II.
SUMMARY
OF SQN DESIGN BASIS VERIFICATION EFFORTS In the paragraph following NRC information requests (a) through (e) in the October 9, 1996 10 CFR 50.54(f) request, the NRC asked for information regarding licensee design review or reconstitution programs.
Provided below is a summary of principal efforts TVA has already undertaken in this area.
This summary is referenced and augmented, as applicable, in TVA's specific responses to requests (a) through (e).
A.
Backaround Sequoyah Units 1 and 2 were voluntarily shut down by TVA in August 1985 because of questions about the environmental qualification of electrical equipment.
On September 17, 1985, NRC requested in a letter to TVA, pursuant to 10 CFR 50.54(f), that TVA describe the corrective actions which-would be completed prior to restart of any of the TVA operating facilitieg and provide a schedule for longer term corrective actions.
TVA submitted and subsequently updated the Corporate Nuclear Performance Plan (CNPP) and the Sequoyah Nuclear Performance Plan'(SNPP), which contained TVA's corrective action plans.
The CNPP addressed concerns with TVA's corporate management. The SNPP addressed SQN site specific issues, with emphasis on the actions required to start the units.
The SNPP was originally issued in November of 1985 and was last revised in May 1988.2 TVA provided separate submittals that contained additional details regarding the proposed corrective action programs.
B.
Desian Baseline Verification Procram (DBVP)
As part of TVA's response to the NRC's letter of September 17, TVA began a DBVP to reestablish confidence in the accuracy of the existing design basis.
The design basis which resulted from that effc rt has been maintained by the programs and processes described im response to NRC Request (a).
TVA submitted additional information and supplemental information to NRC related to the SQN DBVP on 1
NRC letter from William'J. Direts, Executive Director for operations, to Charles Dean, Chairman - TVA Board of Directors, dated September 17, 1985, regarding a Request for Information, Pursuant to 10 CFR 50.54(f), for TVA's corrective Actions to Address concerns Identified During the Fifth Systematic Assessment of Licensee Performance Report.
2 TVA letter to NRC, dated May 9, 1988, Sequoyah Nuclear Plant -
Revision 3 to the Sequoyah Nuclear Performance Plan.
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December 11, 1986, and December 31, 1986, respectively.3
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The DBVP was established to resolve several problems related to design control that had occurred at SQN.
These problems i
were:
The lack of centralized, established lines of I
i authority, responsibility, and accountability for t
i performance of design control functions needed to
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ensure that design integrity is maintained.
Absence of a centralized design input / design basis, inadequate. engineering evaluations, and a lack of j
detail in design output.
Design and modification control methods that did not l'
provide sufficient coordination among groups to ensure accurate documentation of plant configuration and performance of effective safety evaluations.
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The objective of the DBVP was to re-establish the design j
j basis and evaluate the plant configuration.
Thr er.sent (
j elements of the program were as follows:
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Re-establish the design basis and develop design bat...-
documentation that contains the design basis and licensing commitments (e.g., design criteria, calculations, and drawings).
These documents are reviewed when design changes are made and updated accordingly.
i i
Establish and implement an effective design control process.
l Verify the adequacy of the plant modifications made i
since the operating license was issued for systems, or portions thereof, necessary to mitigate design basis accidents and safely shut down the plant.
Verify functional configurations are supported by 1
engineering documentation.
l Verify that past modifications are supported by l
i engineering analysis, documentation, and appropriate j
tests.
'3 TVA letter to NRC, dated December 11, 1986, sequoyah Nuclear Plant - Additional Information on sequoyah Design Baseline and Verification Program.
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TVA letter to NRC, dated December 31, 1986, Sequoyah Nuclear Plant - Supplemental Information to NRC for Design Baseline and
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Verification Program.
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Verify that past modifications _are made in conformance with licensing commitments and design requirements.
Design basis documents verified or developed during the DBVP included system design criteria, system evaluation reports, j
supporting calculations and drawings.
]
TVA developed the SQN DBVP prior to the issuance of 1
NUMARC 90-12, Design Basis Program Guidelines.
However, TVA subsequently compared the NUMARC guidance with the
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corresponding TVA' programs (e.g. DBVP and corrective action progrc.m) and determined that the TVA programs meet or exceed i
the recommendations contained in the NUMARC guidance i
regarding the scope of design basis documents, the process for evaluating and reporting discrepancies, design basis i
document validation, management and control, and the
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integration of the design basis program with configuration i
management and design control.
C.
DBVP InDlsmentatiOB The DBVP was implemented in two phases for Units 1 and 2 at SQN:
TVA completed Phase I for Units 1 and 2 in 1988 before the restart of each unit.
This phase included the evaluation of those systems, or portions thereof, that are required to mitigate FSAR Chapter 15 Design Basis Accidents, to prevent offsite radiation dosage (caused by FSAR Chapter 15 Design Basis Accidents).in excess of 10 CFR Part 100 requirements, or if they were required for safe shutdown.
Specifically, the Phase I DBVP included those systems or portions of systems required for reactor pressure vessel integrity, containment integrity, core cooling, decay heat removal, and reactivity control.
TVA completed Phase II in December 1989, for Units 1 and 2 at SQN.
Phase II extended the DBVP to include system verifications that were not required for restart for Phase I systems and additional safety-related systems not included in Phase I.
The systems evaluated in Phase II included the residual heat removal system, fuel handling and storage system, and flood mode boration system.
In addition, portions of the reactor coolant system, waste disposal ~
system, spent fuel pool cooling system, and radiation monitoring system that were not included in the Phase I effort were evaluated as part of the Phase II effort.
The Phase II scope for the DBVP was submitted to NRC in a bf 5,
in May 1987 and subsequently revised in October 1988 5 TVA letter to NRC, dated May 12, 1987, Sequoyah Nuclear Plant -
Poetrostart Scope and Schedule for the Design Baseline and Verification Program.
5
Prior to completion of Phase I and restart of SQN Units 1 and 2, NRC reviewed the adequacy of SQN's program as, well as
.the implementation of the program.for SQN Units 1 and 7 8,9,10,11,12 In addition, a special NRC DBVP inspection 2.
s was conducted in June and July of 1987 to gsess the implementation and completion of the DBVP.
These inspections and safety evaluation reports concluded that the DBVP effort was adequate and NRC noted the following:
The DBVP was generally conducted in accordance with the program plan, Implementation by both DBVP personnel and the o
Engineering Assurance oversight group appears to have been adequate in most instances sampled by the inspection team, and The extent and depth of inquiry evidenced by the DBVP e
was satisfactory.
TVA responded to the issues raised by the DBVP inspection prior to receipt of the associated team in October 19g, In' this response TVA addressed each inspection report.
NRC item identified in the inspection team exit. The concerns involved applicability of generic reviews for a problem identification report, missing calculations for solenoid valve support variances, adequacy of radiation monitor corrections, controls for punchlist items, and separation of auxiliary control air lines. For each item, 6 TVA letter to NRc, dated october 20, 1988, sequoyah Nuclear Plant - Postrestart scope and Schedule for the Design Baseline Verification Program.
7 NRC letter to TVA, dated May 18, 1988, Safety Evaluation Report on Tennessee Valley Authority sequoyah. Nuclear Performance Plan - NUREG 1232, volume 2.
8 safety Evaluation Report on Tennessee Valley Authority sequoyah Nuclear Performance Plan, Sequoyah Unit 1 Restart - NUREG 1232, Volume 2, Supplement 1, January 1989.
NRC letter to TVA, dated September 15, 1986, Inspection Report Nos.50-327/328 86-38.
10 NRC letter to TVA, dated November 15, 1986, Inspection Report Nos. 50-327/328 86-45.
- 11. NRC letter to TVA, dated February 8, 1987, Inspection Report Mos. 50- 327/328 86-55.
12 NRC letter to TVA, dated June 4, 1987, Inspection Report Nos.
50-327/328 87-14.
13 NRC letter to TVA, dated December 3, 1987, Inspection Report Nos. 50-327/328 87-31.
14 TVA letter to NRC, dated october 27, 1987, Sequoyah Nuclear Plant.- Response to Findings Identified During the Final NRC Inspection of the Design Baseline and Verification Program.
6
TVA described the actions that had been or would be taken to resolve the concerns.
D.
Determination Of Desian Basis Reauirements The design basis was established for Units 1 and 2 using the
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following four steps:
l Step 1 -
Review of Design Basis Documents.and Re-establishment of the Design Basis Step 2 -
Defining the As-constructed Configuration of the Systems and Components Within the DBVP Boundary Step 3 -
Reconciliation of the As-constructed Configuration with the Re-established Design Basis Step 4 -
Issuance of Required Documentation and Corrective Actions Details of each of these steps are provided below:
step 1 -
Review of Design Basis Documents and Re-establishment of the Design Basis:
The design basis input for both units was established-through the creation of a commitments / requirements (C/R) database, design criteria documents, a set of essential calculations within the scope of the DBVP, and functional testing requirements.
In general terms, this was accomplished in the following manner:
The licensing commitments and design requirements necessary to achieve safe shutdown were documented in a C/R database for use in establishing the design criteria documents (DCDs).
DCDs define the engineering requirements necessary to meet the plants' design basis.
Items for the data base included requirements from the FSAR Chapter 15 accidents, abnormal operaticnal transients, special events (e.g.,
evacuation of_the control room), and ext?rnal events (e.g., flooding, tornado, and earthquake) from which the plant must be able to achieve safe shutdown.
The required systems and portions of systems were subsequently identified and evaluated in system evaluation reports.
Essential calculations are those which address plant systems or features whose failure could result in the loss of reactor coolant system integrity, loss of ability to achieve safe shutdown, or a release of radioactive material offsite in excess of the 10 CFR 100 guidelines.
The essential calculationy needed to verify the adequacy of the design were given 7
i i
a technical adequacy review by a third party to ensure that they supported the units' design basis.. The 1
calculations which failed to meet minimum technical adequacy requirements were revised accordingly (or issued, if no calculation existed).
Testing requirements for the verification of system t
j capability were developed, where necessary, to ensure the system configuration satisfied the design basis.
These test requirements were included in the Restart l
Testing Program.
When applicable, these requirements j
were incorporated into testing procedures that periodically verify the operability of systems and l
components.
step 2 -
Defining the As-constructed l
Configuration of the systems and Components Within DBVP Boundary:
The plant configuration was established through functional walkdowns, which compared the plant configuration l
with the as-constructed plant drawings, and through verification of electrical diagrams by walkdowns and functional testing. This was accomplished in the following manner:
Required flow, control, schematic and elementary, and j-single line diagrams were walked down to verify the functional configuration of the system within_the safe
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shutdown boundaries.
The walkdowns verified the arrangement of components, type of components, and i
nameplate data on the components.
The walkdown j
information was documented on drawings to reflect the as-built configuration.
These drawings were reviewed for conformance to the design basis and, following i
reconciliation of discrepancies, were then issued as l
configuration control drawings (CCDs).
The functional configuration of electrical systems and l
l, electrical aspects of mechanical systems were verified in the following manner: 1) Schematic and elementary diagrams were verified by tests and review of existing j
documents, or 2) Single lines were verified by physical l
walkdown and functional testing.
After the functional i
configurations of the electrical diagrams were verified 1
and open items identified by the DBVP were resolved, i
CCDs were issued to document the as-built plant configuration.
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In addition, the results of walkdowns from other j
programs were reviewed (e.g., environmental qualification,.the seismic qualification of large bore 4
piping and pipe supports, field verified cable routing i
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4 data from TVA's cable ampacity and Appendix R program, walkdowns and evaluation of Class 1E low voltage power, j
control and instrumentation cables and safety-related l
medium voltage cables, and the walkdown inspections of l
unqualified coating).
This information was assessed l
and used whenever appropriate in determining the i
functional configuration of the plant.- Results of the restart testing program were used as noted in f
verification of electrical diagrams above.
l Step 3 -
Reconciliation of the As-constructed
]
Configuration with the Re-established Design Basis:
The plant configuration was evaluated against the design basis to verify that the plant was in conformance with safe shutdown requirements.
The evaluation of each plant system (within the DBVP program) was' documented in a System l
Evaluation Report (SYSTER).
This was accomplished in the following manner:
Plant drawings were reviewed-to ensure that they i
accurately depicted the DBVP boundary system functions.
l Differences between the as-built drawings and the as-designed drawings were evaluated and reconciled to i
generate configuration Control Drawings.
Acceptable differences were incorporated into the design basis of the plant and discrepancies were punch-listed and
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tracked until closure.
a.
Corrective action program documents were evaluated to l
e ensure that the corrective actions were in accordance
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with the plants' design basis.
This included closed j
corrective action documents not supported by
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engineering or design change notices.
The results of this review were included in the SYSTER and corrective action was initiated as necessary.
The work performed by other restart programs (environmental qualification, cable tray support i
analysis, design calculation review, alternately analyzed piping and supports, main steam line break j
temperature issues, fire protection-Appendix R, plant i
welding program,' sense line issues, wall thinning assessment program, restart test program, component and l
piece part qualification, electrical issues, containment isolation design review, miscellaneous program, technical specifications, training, corrective
.1 action, quality assurance, employee concerns, new design control program, surveillance instruction procedures program, maintenance, operational readiness j
review, radiological control, operability review, j-functional testing, Sequoyah activities list, procurement, and control room design review) was reviewed to determine the portion that could be used to n
9
satisfy DBVP requirements.
Where appropriate, credit was taken for this work in' evaluating the system configuration.
Change documentation was reviewed on those systems and portions of systems required for safe shutdown to ensure that the changes required to meet the design basis were implemented prior to restart.
Changes that were not implemented or partially implemented were evaluated for acceptability based on their status at the time of restart.
Electrical test reviews were performed for electrical
-and instrumentation control system within the scope of the DBVP to ensure that adequate test documentation
+
existed to demonstrate the functional capability of safety-related systems.
Where adequate test documentation did not exist, a one-time restart test to demonstrate functional capability was performed prior to restart.
l step 4 -
Issuance of Required Documentation and Corrective Actions:
The outputs of the DBVP consisted of j
design criteria' documents, configuration control' drawings, essential calculation technical adequacy review revisions, system evaluation reports, and documentation required to resolve open items for systems within the DBVP boundary.
A SYSTER was prepared for each system within the DBVP to document the evaluations performed as discussed in the above sections.
Open items, including verifying assumptions in essential calculations, reviewing test results, and other l
pre-restart items were punch-listed and tracked until closure.
E.
Independent Reviews In addition, TVA's Engineering Assurance performed independent reviews of the DBVP on a sampling basis using a team of experienced technical personnel.
The objectives of the reviews were to:
Confirm and validate that engineering activities were o
conducted in accordance with the approved program plan and procedures, i
confirm technical adequacy of system evaluations and e
completeness / correctness of supporting documentation.
l J
Verify that corrective actions resulting from these evaluations had been documented and properly implemented.
10
This review provided added assurance that the engineering activities associated with the program wer<t conducted in a technically adequnte manner and in accordance with the written procedures prepared specifically for the DBVP effort.
l 1
The methodology used to establish the as-built plant I
Unit 2.
However, the Unit 1 DBVP incorporated lessons configurations for Unit 1 was the same as that used for learned from the Unit'2 program.
In summary, TVA committed to perform a design basis i
verification in response to a NRC request for information l
addressed to TVA's Chairman of the Board.
TVA performed the i
design basis verification for both units at SQN.
As part of l
this effort, TVA also upgraded its design control process to
{
ensure that changes to the facility are evaluated against, i
l and appropriately reflected in, design basis documents.
F.
Restart Test Program Also important to the information requested in NRC's i
October 9, 1996, 10 CFR 50.54 (f) letter, is the Restart Test Program (RTP) developed by TVA in 1987 to ensure that system functions are adequately' proven to operate (perform) correpp3gpriortotherestartoftheSequoyahunitsin 1988.
The program addressed systems or subsystems which were part of the DBVP effort.
j The major elements within the scope of the RTP were as l
follows:
t Identify the " designed" functions for normal and off-normal conditions of the equipment / systems covered by i
che RTP.
[
i Review these functions against existing test documents i
to ensure they have been adequately tested.
Review these functions and associated test documents l
against applicable on-going restart programs to ensure j
l that associated test results are not compromised.
This i
identification and review process was documented on a i
j functional review matrix and incorporated into a
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functional analysis package.
i 15 TVA letter to NRC dated May 26, 1987, Sequoyah Nuclear Plant -
IWetad: Test Program.
i TVA letter to NRC dated July 6, 1987, Sequoyah Nuclear Plant -
IWetart Test Program.
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Develop additional testing by upgrading existing test documents or generating new tesh documents to prove L
correct operation of functions that are determined not l
to be adequately covered by existing test documents.
Review and approve the above process for technical adequacy and scope by a joint test group.
Implement-RTP required tests.
j i
l Evaluate RTP required " test results" both previously completed and those implemented during the restart effort, for acceptability.
i The RTP was reviewed by NRC as documented in NUREG-1232, Volume 2 (Section 4.9).
The program was found to be acceptable for ensuring the functional integrity of SQN l
safety-related systems.
i G.
Summarv The discussions in this section have described the DBVP and
'related efforts associated with the 1988 restart of both SQN units which effort established an accurate design basis for l
those units.
The following responses to the NRC's specific questions also provide the basis for concluding that TVA has~
adequately maintained control over design basis requirements since that time.
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III.
SPECIFIC NRC REQUESTED INFORMATION i
A.
Request fa)
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Description of engineering design and configuration control j
processes, including those that implement 10 CFR 50.59, j
10 CFR 50.71(a), and Appendix B to 10 CFR Part 50.
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TTA Response to Request (a)
TVA has several interconnecting design and configuration l
control programs and processes, which also address
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10 CFR 50.59, 10 CFR 50.71(a), and Appendix B to 10 CFR Part 50.
These programs and processes include or ara l
l augmented with training, self-checking, and line and
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independent' organization reviews.
The programs at all sites j
contain, as a minimum, the essential elements described in j
this response, but there are minor implementation j
differences between the sites to address specific issues.
Additionally, although these programs have evolved with l
time, the same. basic program features have been in use since thd 1988 DBVP.
Use of these programs, coupled with various oversight activities and other programmatic controls that are described later in this response, provides TVA l
confidence that the SQN design basis has been properly maintained since 1988.
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1.
Desian had Configuration Controls TVA's i
configuration management program is an integrated process designed to ensure that plant SSCs conform to approved design requirements, including design basis, and that the l
plant's physical and functional characteristics are accurately reflected in design basis and other plant documents.
Plant configuration is controlled throughout the life of the plant by the. identification and documentation of i
design requirements and through' procedures which ensure that the design is implemented properly.
Three primary processes are used, as applicable, to implement configuration i
management as applied to changes to SSCs.
These processes i
are:
The plant modifications and design control process, 4
which is the respo Materials Manager;ppibility of the Site Engineering and
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5 17 Nuclear Power Standard 9.3, Plant Modifications and Design j.
Change control.
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t-The nuclear fuel / core component design change control process, which is the gsponsibility of the Corporate Nuclear Fuele Manager; and The temporary alterations control program, which is also the responsib MaterialsManager.gityoftheSiteEngineeringand Plant modifications are implemented using the design change control process, which, in general, is as follows:
a.
Desian Chance Control Process:
A Design Change Notice (DCN) package must be developed. This not only includes design changes to safety-related SSCs, bu is also used for design changes to nonsafety-related SSCs.pg DCN packages are required for design changes that involve plant i
modifications, document-only changes, generic system / component changes, or other changes to an operating nuclear power plant that also involve a design output.
l document change. DCN packages may be used to update design basis documents.
The DCN package provides a basis for the change including references to supporting analyses with new or revised calculations that support the change.
DCN packages are developed from a range of inputs including i
Technical Specifications, design criteria, applicable regulatory requirements, industry and TVA codes and standards, and other similar design considerations in accordance with administrative procedures.
DCN packages i
include 10 CFR 50.59 reviews as required.
Other key DCN j
process features include the following.
Implementation of DCNs (e.g., using work orders) includes installation instructions or references to those instructions.
DCN packages also specify the required post-modification testing necessary to ensure design basis requirements are met.
The preparation and approval of these packages include appropriate multidiscipline and independent reviews and reviews by affected organizations, as required.
For example:
=> An authorized Nuclear Inspector / Authorized Nuclear Inservice Inspector (ANI/ANII) reviews work orders (WOs) which contain work steps affecting ASME Section XI components and/or systems; f
Nuclear Power Standard 9.2, Nuclear Fuel / Core Component Design 18 Change Control.
II Nuclear Power Standard 12.4, Temporary Alterations Control An exclusion list may be established to identify site features l
that are not subject to configuration management control. The list can l
include only SScs that are not quality related and are not described in the FSAR.
The list must be approved by the site Vice President.
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=> RADCON reviews DCNs which involve work within the Radiologically Controlled Area.
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=> An Independent Qualified Reviewer then reviews the l
l' WO prior to its issuance for implementation.
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The DCN process also includes a Return to operation l
l evaluation that is required to be completed before the l
turnover of a modified SSC to plant operations.
This q
l process ensures that operations, maintenance and testing procedures have been updated, that training required to support operability has been completed, and i
i that control room drawings have been revised, i-1 4
l The DCN process includes the updating of the design
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basis documents that were validated by the DBVP (i.e.,
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design criteria, drawings, and supporting 1
j calculations), and any required UFSAR changes.
The
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l package is archived for future reference.
TVA has established a controlled data base which i
contains information relating to the design and I
operational characteristics of as-installed plant l
components.
Processes are in place which update, l
maintain, and control key data to track components in j
the plant by location, unique identifier, description, i
type, manufacturer, etc.
As plant design changes occur and components are replaced, the design control process
[
requires that the data base be updated to reflect the change.
In addition, efforts are planned at each site i
to enhance completeness and correctness of this key i
data.
I Each of these major points in the DCN process discussed l
above includes coordination between affected organizations, i
documentation of the activities performed, as well as j
documentation of the overall change being made.
Management j
i involvement is also a part of these activities.
This i
j includes approval of the request for a change, approval of the engineering work provided in the DCN package, and approval to implement the change.
The work packages provide detailed instructions with self-checking sign-offs in the installation process.
Further, the work package also j
specifies Quality Control hold points for inspection of j
critical activities before installation can proceed.
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TVA assures the'DCN process is followed and is effective.
Engineering personnel who independently perform or technically review safety-related and quality-related design change activities are required to receive indoctrination and 15
training based on Institute of Nuclear Power Operations Academy 91,-017 guidelines.
This training is routinely updated to incorporate lessons learned.
The implementation of the above described controls provide y
reasonable assurance that the design is maintained, however, some specific aspects of the design have been determined to be less than adequately controlled.
For example, the implementation of design changes to abandoned equipment in the plant has been identified as a concerri in the SQN corrective action program.
This concern has identified components such as radittion monitors, boric acid evaporators, auxiliary essential raw cooling water components and other equipment, that are no longer used or intended to be used and requires design changes to properly abandon.
Corrective actions have been developed that will implement design changes to provide appropriate isolation of the abandoned equipment, to complete modifications to d
abandon the equipment, or to restore the equipment to an operable status.
Some of the abandoned equipment identified was determined to be in an acceptable configuration such that it could be left as is or.only required documentation changes to provide an acceptable resolution.
These corrective actions will ensure that abandoned equipment in the plant is properly reflected on design documents and are l
appropriately included in the design control process.
l b.
Fuel / Core component Chance Control TVA controls nuclear fuel / core component design using a similar process.
Generally:
A Core Component Design Change Request (CCDCR) must be developed for modifications to nuclear fuel assemblies i
and other core components.
This includes 10 CFR 50.59 reviews, as required.
CCDCR's are reviewed by each affected plant organization.
This review includes the PORC.
Modifications made on-site to core components are l
completed in accordance with the design change process used for other plant changes described above.
This includes detailed installation instructions, where applicable, and preparation of work orders (Wos).
In addition, the Plant Manager's approval is required before beginning work at the site.
The CCDCR closure process includes, as applicable, the updating of the design basis documents, maintenance, testing and operating procedures, and the UFSAR.
The CCDCR documentation is archived for future reference.
16
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c.
Temoorary Alteration control:
Temporary plant l
alterations is the third process used to implement changes under the configuration management system.
A temporary alteration is normally used to maintain equipment in service using an approved, alternate means until the equipment can l
j be returned to its permanent configuration.
It should be i
noted that_ alterations may be made without using the temporary plant alteration process when the components or j
systems are taken out of service (blocked, tagged, or otherwise inhibited or tripped) for troubleshooting, calibration, modification, or maintenance using an approved procedure and the operability of the affected component or j
system will be verified.by testing prior to returning it to j
service.
Additionally, temporary plant alterations may be j
l made in accordance with approved plant procedures.
If this j
is done, and the temporary conditions are required to remain
[
in place after closure of the approved procedure, the i
temporary plant alteration process is required to be followed.
The general process for implementing a temporary l
i alteration is as follows:
I A Temporary Alteration Control Form (TACF) is developed.
The TACF describe the alteration (e.g.,
Wire lifts, Jumpers, Inhibits, and Temporary
]
Connections), its effects on equipment and functions, i
and its location.
Additionally a Safety Assessment I
(SA), and as necessary, a safety Evaluation (SE) is performed as required by TVA's 10 CFR 50.59 process
{.
(described below).
TVA requires verification of both I
the installation and removal of the temporary alteration.
The plant manager approves temporary alterations.
i Depending on the activity performed, the TACF must be i
approved by the Shift Manager or Unit Supervisor on the affected unit.
In addition, the PORC reviews temporary j
alterations on quality-related components.
i Affected procedures and control room drawings are e
j modified as required to reflect temporary alterations.
l When a unit startup is in progress, an Operations representative (Senior. Reactor Operator (SRo), Shift Manager, or Unit Supervisor) is required to review outstanding TACFs on the unit to determine if the TACF has any restriction with respect to mode changes and power changes.
I j
When a unit is in operation, a periodic review of TACFs e
is performed to determine continued need and to identify any administrative errors in the TACF records.
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i The original TACFs, associated 10 CFR 50.59 review, TACF Audit Sheets, and attached reference drawings for temporary alterations that are installed in the plant are maintained by the Shift Manager.
The closed out TACFs are subsequently archived for future reference.
{
The list of open temporary alterations and the schedule for l
l their resolution is reviewed routinely by senior management.
d.
Desian Process Performance Monitorina Site l
Engineering routinely monitors indicators of the health of l
j the design control process.
Monthly, and in'some cases, l
weekly data is reviewed to track status and cycle time for various engineering deliverables.
Items tracked include DCN i
f' closures, drawing updates, open corrective action program 1
documents, and open TACF's.
Managers use this performance i
monitoring information to focus on and improve process or performance weaknesses.
l 2.
Activities That Maintain Desian confieuration TVA has several layers of administrative 1y controlled programs and practices to ensure that the plant configuration, which j
is primarily controlled by the three programs / processes described above, is mainnained in accordance with the design j
basis established in the design criteria, calculations, and drawings.
Examples of programs and practices that maintain 4
plant configuration durir.g operations and after maintenance or modifications activities are described below.
a.
system Line-un controls:
System line-up is controlled through the use of Equipment Alignment checklists.
Control is initially established through a thorough documented walkdown and alignment of system i
components in the proper configuration.
Verification of this type takes place in situations such as when major evolutions are performed that involve several manipulations to a system's configuration.
The Unit Supervi9or/SRO I
ansures that system line-ups are controlled by directing required changes to the system configuration, ensuring the Configuration Log book is properly maintained, reviewing the Configuration Log book at least once per shift for proper d
usage, and supervising the overall use of the unit's Configuration Log book to ensure status control is 4 -
maintained.
The log book is normally used to document component manipulations during maintenance activities and to-document special system alignments.
)
Quality Assurance periodically verifies that system i
configurations are controlled in accordance with procedures through assessments, scheduled audits, and routine observations.
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b.
Overall control of the operation of Plant plant is directed by the. shift Manager.gpl operation of the Equipments During operations, the over j
The Unit i
operator, Unit Supervisor, and the Shift Manager are j
informed and aware of significant activities affecting plant J
l equipment.
However, activities that are unlikely to affect j
j safety, regulatory requirements, or operating capability (e.g., pumping sumps) may be performed without informing the j
control room.
These activities are coordinated by
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operations personnel outside the control room.
i Additionally, some sample and instrument loop isolation i
valves are configured by other plant organizations and
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controlled by appropriate procedures.
Operator tours are l
conducted by a nonlicensed operator and reviewed by a licensed operator supervisor each shift to maintain l
cognizance of equipment status.
Equipment deficiencies are j
documented and assessed to determine if compensatory actions i
are required by operations personnel.
These compensatory l
actions required by operations personnel are commonly l
referred to as " operator workarounds."
The Operations Superintendent or his designes evaluates the aggregate
.j impact of these identified "workarounds" to ensure that safety and overall operational efficiency is not compromised.
Short and long term corrective actions are prioritized, scheduled, and resolution of these workarounds is coordinated with Operations management and supporting organizations.
c.
control of rauinment Durincr Maintenance and Modifications:
Administrative controls are in place for
)
initiating, planning, performing, completing, and tracking work necessary for both the resolution of operator workarounds that do not involve a design change and performing preventative and corrective maintenance.22,23 These administrative controls require that WRs or WOs be prepared and that any necessary short term configuration changes (e.g., jumpers, wire lifts, temporary instrument settings, unbolting flanges, temporary connections) and status control changes (e.g., repositioning of valves, breakers, or switches) be listed in a log or approved procedures.
Equipment clearances are required before any maintenance is performed on equipment where the unexpected energizing, startup, or relgpse of stored energy could cause injury or equipment damage.
Within the work scheduling process, a Work Week Manager develops a detailed work 21 Nuclear Power standard 12.1, conduct of operations.
22 Nuclear Power standard 6.2, Maintenance Management System.
23 TVA allows certain minor maintenance activities that are comumensurate with craft qualifications and require little coordination to be performed under less restrictive controls than those described in this section.
24~ Nuclear Power Standard 12.3, zquipment Clearance Procedure.
19
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schedule that integrates the activities for each week.
Licensed Operators, System Engineers and others participate in this schedule development to evaluate the impact on plant operations and ensure proper coordination.
Operations is notified before the start of maintenance or modification activities for evaluation of planned configuration changes and so that control of the status of the equipment can be transferred to the maintenance or modifications personnel.
Signatures are required to document that each of the individual configuration and status control changes are made and another signature is required to document when the individual configuration and status control changes are returned to their original position.
A complete system status verification may be performed when major evolutions involving several manipulations are performed.
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operations is notified of the completion of the maintenance task or modification activities and status control of the equipment is returned to them.
d.
Post-modification / Post-maintenance Testina:
Post-modification tests, which are specified as part of the configuration control process, or post-maintenance tests, which are specified as part of the maintyp3gce control processes, are conducted to ensure that:
Equipment performs its intended function following maintenance or modification activities; The original deficiency (if any) has been corrected; i
and A new deficiency has not been created by the maintenance or modification activity.
Quality Assurance periodically evaluates the implementation I
of the post modification / maintenance test (PMT) program through audits and assessments.
These verifications ensure that the plant PMT program includes appropriate plant equipment and verifies equipment will perform its intended i
function.
Some PMTs are verified through Quality Control inspections, such as piping systems that are verified through hydrostatic testing.
a.
Control of Replacement Parts Durina Maintenance:
Configuration control is maintained during the maintenance process when worn or damaged equipment requires 25 Nuclear Power Standard 6.1, Conduct of Maintenance.
26 Nuclear Power Standard 6.2, Maintenance Management System.
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i replacement.
When replacement parts are required, the requestor must provide sufficient information to determine i
thenecessarytegnicalandqualityrequirementsforthe requested items.
1
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Replacement guality-related materials receive a receipt inspection.2 In addition to verifying that the specified f
~
i technical and quality requirements are met, specific j
guidance is provided for identification of transport damage, l
counterfeit and substandard parts.
Requirements for the handling and storage of spara parts are provided to ensure j
that these items are handled, stored, and shipped in a j
manner to prevent deter g tion, contamination, damage, or loss of identification l
Quality Assurance performs regularly scheduled audits of the replacement parts programs to verify that these programs are being adequately implemented.
Programs audited include l
procurement, receipt inspection, storage, handling, shipping and issue, and return. -In addition, audits of supplier's quality programs are performed on a three year frequency to ensure that these programs are being adequately implemented.
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f.
Independent /Second-check / concurrent j'
Verification:
In addition to these configuration controls, i
the operations Manager is responsible for designating those verification.g components requiring independent i
systems and/o The responsible managers are responsible for determining the appropriate type of verification 3
j required for plant instructions and other work documents.
j These types of verification requirements are reflected in site operations, maintenance and testing procedures, i
instructions, and work documents.
For example, breakers, j
. valves, and other components in designated systems are required to be independently, second-check, or concurrently verified to be in their correct position or condition after they'are manipulated for operation, maintenance, modification, or testing activities.
This provides additional assurance that the plant configuration is i
maintained in accordance with the design documents.
g.
Ans_ejas to Desian and Licensina Basis Information The Nuclear Quality Assurance Plan (NQAP)
]
27 Nuclear Power Standard 10.1, Procurement of Material and j
- services, j
28 Nuclear Power Standard 10.2, Material Receipt and Inspection.
j 29 Nuclear Power Standard 10.3, Handling and Storage of Materials i
and Spare Parts.
30 Nuclear Power Standard 10.4, Material Issue, Control and l
Return.
31 Site Standard Practice 12.6, Equipment Status Verification and Checking Program.
21
4 requires that for activities affecting quality, measures shall be established to ensure that documents prescribing theactivgyaremadeavailabletopersonnelperformingthe activity.
The TVA Document Control and Records Management (DCRM) program defines the process for the control of and access to drawings, specifications, design criteria, and other documents related to design basis.
Access to principal elements of plant licensing basis, such as the FSAR, Technical Specifications, and correspondence submitted to NRC, is also controlled through this program.
l Design and licensing basis documents are controlled to ensure that the latest version is used in performing activities affecting safety.
3.
10 CFR 50.59 Safety Assessments / Safety Evaluations:
The 10 CFR 50.59 process g controlled by a Nuclear Power administrative procedure.
The procedure addresses changes to the facility or prgedures described in the Safety Analysis Report (SAR) or tests or experiments not described in the PAR to determine if an unreviewed safety question exists.3 The process includes a SA which consists of a determination of the acceptability of a proposed change from a. nuclear safety standpcint, and a screening review to determine if the activity would result in 1) a technical specification change, 2) a change (other than administrative or editorial) to the information presented in the SAR, or 3) if a test or experiment is not described in the SAR.
If the SA indicates that the proposed activity might not be safe, the activity must be modified or canceled.
If the activity is determined to be safe, the process continues to evaluate whether a Technical Specification or SAR change is involved.
If a Technical Specification change is found to be i
necessary, a license amendment is submitted to the NRC for approval.
If it is determined that a change (other than administrative or editorial) is being made to the information presented in the SAR or a proposed test or experiment is not described in the SAR, a SE addressing the questions in 10 CFR 50.59 is prepared.
If it is determined that the proposed chang,e, test, or experiment involves an unreviewed safety question, then the proposed action must be revised, canceled, or reviewed by the NRC prior to J
TVA letter to NRC dated August 31, 1995, TVA Nuclear Quality Assurance (NQA) Plan (TVA-NQA-PLN-89-A) Update - Revision 6.
Nuclear Power Standard 12.13, 10 CFR 50.59 Evaluations of Tests and Experiments.
Changes e 34 This includes the latest updated FSAR, FSAR changes not yet incorporated in the controlled FSAR, and any licensing basis commitments not yet incorporated in the controlled FSAR.
35 For changes to commitments not within the scope of 10 cFR 50.59, TVA has a commitment change process, which is described later in this response.
22
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implementation.
The SA and, if it is required, the SE are prepared as part of the design control or-procedure change
+
process prior to the implementation of-the change or initiation of testing.
The program requires that only qualified personnel prepara and technically review SAs and SEs.
Preparars and reviewers must be formally trained before working on SAs and SEs and they must receive retraining at two-year intervals.
i Typically, the initial training for 10 CFR 50.59 involves a i
two-day classroom instruction with an examination and a 1
l practical exercise in which an actual SA or SE is prepared.
Required retraining consists of classroom training on topics i
related to the SE process.
i Line managers are responsibla for assigning trained and qualified preparars and reviawers for SAs and SEs consistent with the complexity and scope of the proposed activity.
Preparers are required to obtain technical assistance i
outside their immediate area of expertise and responsibility, when it is needed to complete the SA or SE.
[
Preparers also ensure that the SA and/or SE is ' consistent
{
with the UFSAR (including the "Living FSAR"), NRC operating license amendment safety Evaluation Reports (SERs) (including j
supplements), major restart program SERs, and plant Technical
+
Specifications.
The PORC reviews selected SEs as an i
oversight function of the 10 CFR 50.59 activities.
The Nuclear Safety Review Board (NSRB) provides oversight of the i
SA/SE process and periodically assesses the adequacy of the 10 CFR 50.59 Program.
l TVA's 10 CFR 50.59 program has evolved and will continue to evolvr., in consideration of industry and regulatory practices as well as to address performance issues identified by TVA.
As an example, a substantial change to the program was made when NSAC 125 was issued.
Other changes were made to provide appropriate program guidance when new issues have arisen such as evaluating analog to digital control system changes or if internal or external assessments indicate that the program should be enhanced.
4.
Undated Final safety Analysis Resort rio cFR so.71te11:
TVA administratively controls the UFSAR, including how this document is r in accordance with 10 CFR 50.71(e).gyised and updated,.
Changes to the UFSAR are identified during the performance of the SA/SE process (required to comply with 10 CFR 50.59), and during the preparation of design changes or procedural revisions.
In addition, not only can an individual identify the need for a change to the UFSAR, they also have the responsibility to identify any known errors within the UFSAR.
Nuclear Power Standard 4.2, Management of the Final Safety Analysis Report (FSAR).
23
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i Changes to the UFSAR aust be made in accordance with the administratively controlled process which is coordinated by the Nuclear Assurance and Licensing (NAEL) Department.
The i
L procedural requirements for submitting a change to the UFSAR l
include:
2 A UFSAR change form must be completed, which includes specific references to the pages, figures, and tables, i
that require revision.
l The preparer must provide annotated pages, figures, tables, or replacement pages that clearly indicate the requested change.
Identification of the date that the activity addressed by the UFSAR Change Request was implemented (field l
complete or plant approved).
This date is used to i
ensure that the UFSAR is up to date as of a maximum of six months prior to the date of filing the amendment in accordance with 10 CFR 50.71 (e) (4).
l The preparer must also provide the supporting l
justification for the change.
This normally consists of the SA/SE performed in accordance with 10 CFR 50.59.
However, the justification may also be in the form of an NRC Safety Evaluation Report (SER) that addresses l
the subject of the change request, such as the SER from an NRC approved operating license amendment, or justification that the UFSAR Change Request is an administrative change.
In accordance with the administrative controls for the UFSAR change process, the NA&L logs and tracks UFSAR changes and ensures that the organization assigned primary technical i
responsibility for the affected UFSAR section evaluates the proposed UFSAR change.
Approved changes are periodically incorporated into the Living FSAR, so that there is access to the latest FSAR material.
The Living FSAR is a document that compiles approved FSAR changes that have not yet been incorporated into a UFSAR amendment package.
In order to prepare a UFSAR amendment, NAEL consolidates individual changes that have been implemented prior to the UFSAR amendment cutoff date.
NA&L coordinates a multidiscipline review of the UFSAR amendment submittal to NRC with appropriate concurrence in accordance with the administrative controls established for yritten communications between TVA and the NRC.3 Once the UFSAR
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37 Nuclear Power Business Practice, BP-213, Managing TVA's Interface with NRC.
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amendment is approved for submittal to NRC, the controlled copies of the UFSAR are updated in accordance with the administrative controls.
i Assessments of UFSAR accuracy performed in conjunction with
.TVA's most recent periodic amendments at.BFN and SQN identified a number of discrepancies.
None'of these discrepancies resulted in Unreviewed Safety Questions, rendered systems. inoperable, or required modifications to' plant equipment.
Additional details are provided in TVA's response to NRC Request (e).
l l
5.
Yanlementation of Annandix B To 10 CFR Part 50s The TVA NQAP describes the Quality Assurance (QA) Program as required in 10 CFR 50.34, " Contents of Applications; Technical Information" and 10 CFR 50.54, " Conditions of Licenses."
The QA Program described in the NQAP provides control over activities affecting the quality of identified structures, systems, and components to the extent consistent i
with their importance to safety.
The NQAP is referenced in each TVA' Nuclear plant's Safety Analysis Report and has been accepted by the NRC as meeting the requirements of 10 CFR 50, Appendix B.
t
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The NQAP places responsibilities on identified sponsors to develop specific elements of the QA Program, addressing requirements of source requirement documents such as NRC Regulatory Guides and ANSI Standards.
This is accomplished through administrative procedures (e.g., Nuclear Power Standards (STDs), Site Standard Practices (SSPs), Standard i
Programs and Processes (SPPs)) that are normally sponsored by managers who are responsible for designated functional areas.
STDs and SPPs are procedures that define overall program requirements.
Each STD or SPP sponsor is responsible for incorporating into STDs and SPPs, QA and i
other regulatory requirements applicable to the functional i
area.
Site and corporate organizations implement STDs either directly or through lower-level documented procedures or instructions such as SSPs.
SPPs are directly implemented
-at the sites.
The engineering design and configuration control processes described above also incorporate the relevant requirements of Appendix B to 10 CFR Part 50.
For example, the three processes used to modify the plant's configuration satisfy Criterion III, Design Control, and incorporate requirements necessary to ensure that instruction, procedures and drawings are revised prior to closure in accordance with criterion V, Instructions, Procedures, and Drawings.
Procurement requirements necessary to ensure adequate quality of the requested items are specified in order to satisfy Criteria IV, Procurement Document Control; VII, Control of Purchased Material, Equipment, and Services; and 25
VIII, Identification and Control of Materials, Parts, and t
Components.
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The corporate Quality Assurance organization performs audits j
1 to assess the adequacy and effectiveness of the TVA Nuclear QA program.
These audits are performed in accordance with i
written procedures or checklists by qualified and certified i
personnel who have no direct responsibility in the areas being audited.
Audits evaluate a number of quality-related attributes, including:
t Compliance with Technical Specifications and license conditions.
Performance, training and qualifications of the plant staff.
I Effectiveness of actions taken to correct problems with equipment, SSCs or methods of operation that affect nuclear safety.
The performance of activities required to meet the l
criteria of Apperdix B to 10 CFR 50.
l The QA organization audits both TVA Nuclear organizations and contractors and suppliers who provide safety-related j
services or materials.
l Additionally, independent technical reviews are performed by site NA&L to assess activities such as modifications, maintenance, and engineering to verify that these activities are performed safely and correctly.
6.
Commitment Manacement:
A commitment management and tracking process is in place to provide a formal method for controlled by 10 CFR 50.59. ganging commitments not identifying, tracking, and This process incorporates the Nuclear Energy,Inshte M %dehnes for Managing NRC Commitments".
Sources of commitments include Licensee Event Reports, NRC requirements, and NRC safety evaluation reports.
NAEL maintains a commitment tracking system that includes a description of the commitment, responsible organization, and due date.
Changes in the scope or completion date specified in a commitment to NRC must be justified.
As necessary, such changes are submitted to NRC by NAEL, and the commitment tracking system is updated.
Standard Programs and Processes 3.3, NRC Commitment Management.
3' The NEI guidelines were recently included in TVA procedures.
However, TVA participated in the NEI pilot project in 1994 - 1995 and has used the guidelines since that time.
26 i
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NA&L reviews commitment closure documentation to confirm i
that actions taken conform to the original intent of the commitment and that the original concern is satisfied.
NA&L
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a may elect to have closure independently verified for any commitment.
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1 B.
Request (b)
Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing
'l procedures.
j TVA Response to Request (b)
TVA is confident that the SQN design basis requirements have been translated into operating, maintenance, and testing procedures.
TVA's confidence is based on TVA's design change processes, administrative controls for the
)
preparation, review, and approval of new and revised I
procedures, various procedure upgrade.and review efforts, and continuing assessments.
These efforts and assessments involve comparing applicable design and licensing basis I
information to procedures to ensure that the plant is operated safely and in accordance with design and (Icensing requirements.
TVA recognizes that to ensure design basis requirements are accurately translated into procedures, continued vigilance is necessary.
When TVA identifies problems with its procedures, it pursues corrective actions to its procedures and processes, as necessary.
1.
Backgrounds Procedures were developed by TVA for initial SQN operation in the late 1970s and early 1980s.
Many of the surveillance procedures were based on preoperational tests performed prior to initial startup of the units.
This development included the following.
1 Development of Sequoyah (SQN) Operations Procedures for j
initial plant operations involved several inputs including Westinghouse and TVA system descriptions, SQN drawings, SQN Technical Specifications and the SQN I
FSAR.
Once developed, these procedures were verified and validated by plant walk downs, use during pre-operational testing, and use during licensed and nonlicensed operator training.
The initial SQN procedures were developed within the guidance provided by approved plant procedures and processes.
Sequoyah procedures were maintained current with Sequoyah's design after initial plant licensing by following a controlled procedure change process and the design control process.;
Emergency Operating Instructions (EOIs) were revised through an upgrade effort during the 1985 - 1986 time frame to Emergency Operating Procedures (EOPs) 28
[
4 utilizing Westinghouse owners Group recommendations incorporating the Three Mile Island (TMI) lessons learned analysis, the symptom based Emergency Response Guidelines (ERG's), industry experience, setpoints developed and issued by engineering, sequoyah simulator-f verification, physical in-plant verification as appropriate, and Westinghouse review for Emergency Response Guidelines (ERG) alignment and FSAR verification.
r The original surveillance and maintenance instructions were i
developed concurrently with the pre-operational test procedures, using the draft Technical specifications, Westinghouse information, setpoint and scaling information developed by plant engineering groups, and other vendor f
information.
2.
Procedural Controls I
a.
Desian chance Process:
When a design change is made,_TVA's design change process requires that affected procedures be identified and created or revised to reflect the design change.
The design change processes are described above in response to NRC Information Request (a).
These processes provide a line of defense in ensuring that the design is correctly reflected in the applicable operations, maintenance, and testing procedures.
b.
Procedure Generation / Revision Process:
The administrative control processes established for the development, review, approval, and control of the SQN procedures are designad to implement upper tier programmatic and nuclear quality assurance plan requirements.
These controls ensure design basis requirements are adequately i
reviewed for the development of, and incorporation into, site technical and administrative procedures.
Established controls ensure that design basis information is adequately researched byg procedure authors during procedure development.
Programmatic controls ensure that procedures are reviewed by affected organizations (e.g., operations, 1
Engineering, RadChem), independent qualified reviewers (as discussed in response to (a)), quality assurance (if required), and qualified 10 CFR 50.59 reviewers.
In addition to the procedure development and review cycle, some technical procedures receive walkdown' reviews and PORC reviews, as appropriate, prior to approval.
The procedures control program also addresses the revision process.
These controls ensure that procedures are maintained current and reflect plant configuration control j
changes.
Revisions to procedures are processed through the 40 Site Standard Practice, SSP 2.3, Administration of Site Procedures.
29
-wemge w-
-+-----e--
1 I
I i
l same series of controls as newly developed procedures.
5 l
These administrative 1y controlled checks and balances ensure that design basis requirements are correctly translated into operations, testing, and maintenance procedures.
)
c.
Vendor Manual Control Progrant Through the Vendor Manual control Program, TVA incorporates vendor l
procedures.jpnaintoapplicableoperationsandmaintenance recommendat The program establishes and maintains a
{
standardized process which documents the receipt, disposition, deviation, revision and utilization of vendor i
l manuals.
The Procurement Engineering Group (PEG) receives vendor manual information and distributes it for procedural i
l impact review by the applicable organizations (e.g.,
1 Operations and Maintenance).
The applicable organizations review the information and revise their procedures, as 1
i necessary. In this manner, TVA ensures that systems and components are operated, tested, and maintained in accordance with vendor recommendations; thereby ensuring the i
design basis requirements are met.
l
)
3.
Procedure Verificationst in addition to the
)
procedural controls discussed above, TVA has performed j
l several programs and procedure reviews that have verified i
that design basis requirements have been translated into procedures.
As sample of these programs and reviews is discussed below, i
j a.
Procedure Uperade Program:
As explained earlier in this response, TVA re-established the SQN design 4
basis requirements through a comprehensive process and reconciled the as-designed and as-constructed facility as part of the DBVP.
Using the design basis information I
generated from the DBVP, TVA implemented a procedure upgrade program to ensure that design basis requirements were incorporated into operating, maintenance, and testing procedures.
The procedures upgrade program was tasked, in part, with identifying those procedures important to safe i
operation that required' revision or development as a result l
of the DBVP system walkdowns and completed plant i
modifications.
For example, Surveillance Instructions were reviewed to ensure that they accurately reflected Technical Specification requirements in the acceptance criteria.
In i
general, procedures were walked down to verify their practicality and adequacy for implementation.
This effort i
was perforced following the 1985 shutdown of the SQN units.
l Near-term and long-term procedural solutions were developed
{
from the identified deficiencies.
Near-term procedure upgrades were focused on corrections needed to restart and long-term solutions were aimed at a total revamp of the i
U site Standard Practice SSP-2.10, Vendor Manual control.
30
I existing procedure hierarchy.
These pr reviewed and approved by the NRC Staff.pprams were also A procedures' staff was established to correct identified i
procedure deficiencies and to implement the revised upper, tier requirements imposed by the performance plans.
The work performed by this staff ensured that design requirements were correctly translated into procedures.
l This effort is summarized below.
Operations Related Procedures (including operation e
conducted test procedures):
=> SQN Design Baseline Verification
~
Results from the baseline verification were used to revise operations procedures using the normal processes (e.g., drawing deviations, modifications, etc.).
Additionally, procedures upgrade programs I
were initiated during this time frame.
This process involved tabletop review of procedures and appropriate drawings as well as a plant walkdown with the various operating procedures.
These.two efforts resulted in major rewrite of the following:
l Surveillance Instructions (sis)
System operating Instructions
=> Following Design Baseline Verification Following the restart of SQN Unit 1 & 2 in the 1988 time frame, the following procedures upgrades were completed as originally scoped:
EOP program upgrade with the issuance of:
SSP-12.16 - Emeraency Operatina Instruction Control EOI Program Manuals (EPM)
EOP revision 1A implementation EOP revision 1B implementation Abnormal Operating Procedure (AOP) upgrade Annunciator Response (AR) upgrade System Operating Instruction (SOI to SO) upgrade (one remains to be implemented) 42 NUREG-1232, Volume 2, and supplement 1 - Safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nuclear Performance Plan, sections 4.2.2.6 and 4.6.
31
l
}
)
i i
General Operating Instruction (GOI to e
I GO) upgrade i
-l 4
i Each of the upgrades referenced previously was j;
implemented by following a controlled procedure l
change process. Additionally, verification and l
validation processes were used to ensure the i
j adequacy of procedures.
1 Maintenance Related Procedures (including maintenance conducted test procedures):
=> The following actions were undertaken as part of the i
1986, TVA-generated Nuclear Performance Plan:
Enhancement of SQN maintenance
~
instructions (mis) to simplify work packages and to develop new instructions fur the performance of some safety-related component maintenance work.
' Review of Technical Specifications and associated sis to ensure satisfaction of surveillance requirements.
The majority of sis and mis were re-written to ensure unit, train and channel separation.
As part of the enhancement effort, a procedure writers guide was drafted to ensure uniformity in mis.
Key elements to the enhancement effort were defined as follows:
Inclusion of procedural detail to ensure consistent performance and technical adequacy.
Development of mis to address specific tasks that have generic applications.
Human factor considerations were included.
3 Verification and validation of procedures was' performed to reduce errors and improve quality. ~
verification is part of the table top review of the procedure and validation is the walkdown conducted prior to implementation.
The procedures Administration Group developed a checklist as part of an Independent Qualified Reviewer process.
This ensured updated design change information, and other design output j
32
i j.-
1 i
generated from the DBVP effort were incorporated into SQN surveillance and i
maintenance instructions.
This checklist review was adopted in the instruction revision process to ensure the design criteria was l
considered.
l In September, 1991, the enhancement effort had established a program which ensured procedures met'critoria for technical accuracy, l
administrative consistency, incorporation of human factors considerations, standard format i
and organization, and incorporation of
?
craft / performer skill.
Problem Areas Since the procedure enhancement program has been concluded, specific procedure-related issues have beer, identified for which TVA has taken appropriate corrective actions.
However, these issues have not been such as to erode TVA's overall confidence-that the SQN design has been adequately translated into procedures.
Should procedure issues be discovered in the future, TVA will address them in accordance with the corrective action program.
b.-
Operating Experience Program:
TVA's Operating Experience (OE) Program was established po satisfy the Item I.C.S.4 This program requirements of NUREG-0737, evaluates experience. reports received ~from NRC, INPO, nuclear vendors and equipment suppliers, architect / engineers and constructors.
TVA's evaluations of these reports often result in procedure evaluations that utilize industry experiences to ensure design basis information is appropriately reflected in operations, maintenance, or testing procedures.
For example:
In response to NRC Information Notice 94-80, TVA evaluated Electrical Design Guide DG-E7.1.3 to determine compliance at SQN.
This design guide addressed the capabilities of ground detection systems for DC power distribution systems.
TVA's review considered the plant design and procedural controls that were in place and concluded that improvements were needed to ensure adequate ground detection capabilities. 'The improvements included the development of a procedure to remove grounds at a lower j
threshold.
43 Nuclear Power Standard 4.4, Managing the operating Experience Program.
33
1 1
1 TVA evaluated the concerns associated with NRC Information Notice 95-04 regarding excessive cooldown and depressurization of the reactor coolant system.
TVA evaluated the current procedure requirements and in-process changes to SQN emergency operating procedures (EOP).
The procedure changes that were being processed as part of the Revision 1B EOP upgrades satisfied the concern and were consistent with' actions that were discussed in the information notice.
This evaluation resulted in the conclusion that SQN procedures would acceptably address the information notice concern after the procedure revisions have been implemented.
l TVA evaluated plant condition and maintenance instruction adequacy in response to a 10 CFR Part 21 issue regarding defects of the stationary secondary disconnect conductor strips of the ITE SHK Circuit Breakers.
TVA evaluated selected breakers for defects with none being identified.
TVA's evaluation of maintenance instructions indicated that appropriate verifications were included.
However, TVA added the recommendations of the Part 21 notice, to inspect the strips for cracks, to the breaker maintenance instruction as an added enhancement to the procedure.
j Through these types of system specific evaluations of procedures, TVA verifies that plant design basis are correctly reflected in operations, maintenance, and testing procedures.
c.
Generic Regulatory Issues:
TVA often verifies the translation of design basis requirements into operating, j
maintenance, and testing procedures, or enhances those procedures, for specific systems as part of its response to specific regulatory initiatives.
For example, TVA implemented an administratively controlled motor operated valve (MOV) program.
This program was developed as part of the activities to address Generic Letter 89-10 and its supplements.
Procedures were developed for the MOV program to outline the differential pressure requirements and include the design output for the new thrust calculation methodology.
A site procedure was also developed to describe the design and implementation of the program.
This program is a combination of design basis evaluations, proceduralized valve static and dynamic testing, preventative and corrective maintenance, performance trending, and equipment training.
The program and procedures are used to ensure the valves included in the program can be operated as designed under design basis conditions 34
i l
[
d.
Emergency Operating Procedures The SQN EOPs
]
are significant procedures since they direct operator actions to achieve the safe shutdown of the plant after a broad range of accidents and equipment failures.
The
. process for the development and revision of the EOPs ensures 1
that design basis requirements are reflected in the EOPs.
To develop EOPs for SQN, in general terms, TVA translated the design basis into the EOPs in the following manner:
TVA substituted SQN plant specific values for the generic values contained in the Westinghouse owners Group Emergency Response Guidelines (ERGS) by reviewing the SQN design basis.
TVA performed additional calculations to determine SQN specific curves and limits that are used in the EOPs.
TVA developed from its design basis and physical plant characteristics, technical steps not required by the ERGS, but necessary for SQN to successfully mitigate the event.
TVA researched its design basis requirements and deleted ERG steps that did not apply.
Verification and validation of the initial EOPs was performedfnaccordancewiththeEOPadministrative controls.4 TVA has a process to control revisions of the EOPs to ensure design basis requirements are maintained.
This process includes checks and balances designed to verify and validate that the change is consistent.with design basis j
requirements.
This process includes a review of the change by the EOP coordinator to determine whether verification or validation of an CDP change is required and to obtain concurrence from the Operations Superintendent; and a verification and. validation similar to that conducted in the initial EOI development process.
EOPs require Plant Operations Review Committee review and Plant Manger approval.
Significant EOP changes also require Westinghouse Electric Corporation review and concurrence.
4.
Independent Assessments:
In addition to the various programs and procedural reviews TVA has performed, assessments of TVA's procedures challenge TVA's various procedural control processes and the translation of design basis requirements into procedures.
A sample of these assessments is discussed below.
Site Standard Practice, SSP-12.16, Emergency operating Instruction Control.
35
j l
=
a.
ouality Assurance Assessments:
TVA has l-assessed and audited the procedure control and change
-process and procedure implementation on numerous occasions.
i These reviews have determined that overall, procedures and i
procedure implementation are adequate.
Examples are 4
described below:
i j_
An assessment of the independent qualified j
reviewer (IQR) process was performed to determine the effectiveness of this process.- This assessment was I
initiated to verify that the IQR process is capable of l
preventing the recurrence of deficiencies in j
surveillance instructions.
The results of this evaluation were that the IQR training and procedure i
reviews were adequate and being performed as required.
Minor discrepancies were noted regarding attention to j
detail in some IQR reviews and the use of non-intent l
criteria for procedure changes.
I l
TVA conducted a vertical slice review of the main and auxiliary feedwater systems and portions of the safety i-injection system.
This audit included the evaluation of compliance with administrative procedures for activities performed in the engineering, maintenance, i
procurement, storage, and test equipment areas.
The l
audit also evaluated the implementation of design l
criteria through the design change process into plant procedures and instructions.
While isolated discrepancies were identified and corrected in three l
procedures, the overall conclusion was that the incorporation of the design basis into procedures l
adequately maintained the function of the systems evaluated.
b.
MRC Inspections: NRC conducted a Service Water 4
System Operational Performance Inspection (SWSOPI) of the j
i Essential Raw Cooling Water (ERCW) on January 9, 1995. TVA conducted a self-assessment of the ERCW system in 1993 and an assessment in 1994 of service water systems. Over a nine week period, the NRC team examined plant activities in the areas of mechanical and electrical system design, operations, maintenance, surveillance and testing, and quality assurance.
The inspection included an in-depth review of the system's surveillance procedures.
One violation identified during the 1995 SWSOPI inspection cited an example that was based on inadequate procedures and improper procedure implementation.
TVA implemented appropriate corrective actions for this violation which included a change to the design control process to ensure l
that the operations section reviews appropriate civil-related modifications thereby strengthening procedure impact 36
i reviews and providing better design basis control.
Another violation example from this inspection indicated that a procedure used design input to set system performance requirements in place of design output.
TVA corrected this issue by revising affected procedures that referenced design calculations (design input) and provided guidance in the procedure process to emphasize the use of design output in procedures.
These procedure issues were not found by the NRC to render the system inoperable.
The specific issues were addressed including broader programmatic changes necessary to ensure the proper translation of design requirements into procedures.
5.
Summary:
The procedural controls, procedure upgrades, and various procedure reviews described above, provide assurance that design basis requirements are transmitted into operating, maintenance, and testing procedures.
P i
37
C.
Request (c) l Rationale for concluding that SSC configuration and performance are consistent with the design bases.
TVA Response To Roguest (c)
TVA is confident that SQN SSC configuration and performance are consistent with the SQN design basis.
TVA's confidence is based on TVA's configuration control and corrective action processes (as described in response to Requests (a) and (d)), programs that compared SSC configuration to design requirements and tested SSC performance, testing that verifies SSC performance on an ongoing basis, and various reviews and continuing assessments of SSC configuration and performance.
TVA recognizes that it must continue to be self critical in this area to ensure that SSC configuration and performance remain consistent with design basis requirements.
When problems are identified, corrective actions are taken.
1.
Configuration and Performance Controls:
As discussed in response to Request (a), TVA's configuration management program is an integrated process designed to ensure that SSCs conform to design requirements.
Plant configuration is controlled throughout the life of the plant by the identification and documentation of design requirements, and through procedures that ensure that design is implemented properly.
TVA has several layers of administratively controlled procedures and practices in the areas of:
j Design changes, System line-up, Operation, maintenance, modification and testing of plant equipment, Procedure generation or revision, and Vendor Manual Control All of these processes work together to ensure that system, structure, and component configuration and performance are maintained consistent with the SQN design basis.
When problems are identified, they are evaluated and corrected in accordance with TVA's Corrective Action Program.
2.
Verification Proorgani In addition to the various configuration controls discussed above, TVA has implemented several programs that have verified that SSC configuration is consistent with design basis requirements and that SSCs perform in accordance with design basis requirements.
These programs are described below.
38
i i
a.
Reconciliation of Damien Basis to Plants As discussed earlier in this response, TVA established the design basis and reconciled design basis requirements with l
plant configuration as part of the DBVP.
This process involved verifying plant configuration through walkdowns l
(i.e., comparing plant configuration with the as-constructed plant drawings), and verifying electrical diagrams by l
walkdowns and functional testing.
The as-constructed configuration was then reconciled with the re-established design basis, which included performing any necessary testing.
This was accomplished by the performance of a system evaluation.
The system evaluation involved assembling all walkdown and system test review results, implementation documents, and change documentation i
i and then comparing the re-established design basis requirements to the walkdown and test results and the outstanding change documentation.
Differences were l
corrected by modifying plant equipment or revising design documents as appropriate.
In addition, existing corrective I
action documents were evaluated to ensure that corrective actions already implemented were acceptable such that the system safety function was not compromised.
Unacceptable items identified during the system evaluations were j
addressed by the corrective action program.
j b.
Restart Test Program:
TVA performed a j
comprehensive restart test program to ensure that SSCs were capable of meeting their safe shutdown performance requirements confirmed by the DBVP effort.
The restart test program was composed of normal surveillance testing as well as supplemental tests required to address issues identified in the restart review programs and demonstrate that SSCs would perform in accordance with these design requirements.
This testing was performed as part of plant restart efforts.
The NRC reviewed TVA's Restart Test Program and concluded in NUREG-1232, Volume 2, that the implementation of this program would ensure proper verification of the fungpional integrity of the safety systems at Sequoyah Unit 2.
Similarly, the NRC staff determined that the Unit 1 Restart l
TestProgramprovidedthesamelevelofverificap[onand supports the safe return to operation of Unit 1.
NRC letter to TVA, dated May ).8, 1988, Safety Evaluation f
Report on the Tennessee Vallay Authority Sequoyah Nuclear Performance j
Plan - NUREG-1232 volt:me 2, Section 4.9.
46 Safety Evaluation Report on the Tennessee Valley Authority Sequoyah Nuclear Performance Plan - NUREG-1232 volume 2, Supplement 1, j
section 4.9, January 1989.
i 5
39 t
r
i From the restart tests conducted in 1988, TVA verified that these SSCs would perform in accordance with design basis requirements.
l l
3.
Osc Testing
- a. Routine surveillance Testina Surveillance testing is an important tool to demonstrate that SSCs will perform in accordance with their design and licensing requirements and commitments (e.g., Technical Specifications and the Fire. Protection Report).
The Surveillance Program
)
provides the administrative controls for surveillance scheduling, t i
test records.gyting status, evaluation of test results, and Through these controls, TVA ensures that j
testing necessary to demonstrate performance of SSC is i
executed in a rigorous fashion.
(TVA is evaluating applicable surveillance tests in accordance with Generic i
Letter 96-01)
Controls include:
l Plant operational modes for which the surveillance requirements are required to be current to support system / equipment operability are listed.
In the cases where the operational modes are not specifically indicated by the Technical Specifications, the modes which are applicable are determined by the organization j
responsible for the surveillance procedure.
Plant operational modes in which the surveillance can be performed (and therefore, by inference, the modes it can not be performed in) are listed.
The frequency or initiating plant condition / event for each surveillance requirement is listed.
The implementing procedure number for each surveillance requirement is listed.
The organization (s) responsible for performing / preparing each surveillance requirement is listed.
b.
_%BME Code Reruired Performance Monitorina (1)
Inservice Inspection:
10 CFR 50.55a(g) and SQN Technical Specifications both require the establishment and implementation of Inservice Inspection (ISI) requirements (including preservice) in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for components j
i Site Standard Practice, SSP-8.2, surveillance Test Program.
l 47 40
1 1
l (including supports) which are classified as ASME Code l
Class 1, Class 2, and Class 3, or equivalent.
I Administrative controls provide instructions necessary for t
the preparation and implementation of pSME Section XI ISI or l
Preservice Inspection (PSI) programs.4 Implementation of this inspection program and the performance of the l
prescribed testing provide added assurance that safety-related components will perform consistent with their design l
basis requirements.
l 1
l (2)
System Pressure Tests:
Administrative controls also establish the general requirements, responsibilities, and guidelines for preparation, review, i
and performance of system pressure tests (SPTs) to meet the i
requirements of Section XI of the ASME Boiler and Pressure i
Vessel Code, as required by Paragraph 50.55a(g) of 10 and the SPT requirements of Technical Specifications.4pFR 50 j
i i
Implementation of this testing' program provides added l
assurance that the pressure boundaries for safety-related systems will maintain their integrity.
4 j
(3)
Inservice Testina 10 CFR 50.55a(f) and i
Technical Specifications require the establishment and implementation of ASME Section XI Subsections IWP and IWV for Inservice Testing (IST) in accordance with Section XI of 4
j the ASME Boiler and Pressure Vessel Code to verify operational readiness of pumps and valves whose function is i
required for safety.
Administrative controls establish the SQN.ggentation requirements for IST of pumps and valves at l
impl These controls reflect the current "living" IST j
program for SQN, including the applicable Safety Evaluation Reports (SER) and requirements and active changes to the IST j
program in progress.
Implementation of this testing program provides assurance that safety-related pumps and valves will
}
function during anticipated operating, transient, and j
accident conditions.
)
c.
containment Testina Administrative controls program.pishedfortheprimarycontainmentleakratetest areestag This Site Standard Practice provides a list of j
primary containment boundary components and specifies the associated test requireuents for both the integrated l
containment leak rate testing and penetration testing in order to satisfy the requirements of 10 CFR 50, Appendix J.
Implementation of this testing program provides added I
Site Standard Practice, SSP-6.10, ASME Saction XI and Augmented Nondestructive Examinations.
ailta Standard Practice, SSP-8.5, ASME Section XI System Presrure Test Progree.
Site Standard Practice, SSP-8.6, ASME Section XI Inservice 50 Testing 1of Pwqs and Wtives.
Site Standarc Practice, SSP-8.7, Containment Leak Rate Programs.
41
i assurance tha,t primary containment and its penetrations will perform consistent with their design basis requirements.
4.
Continuina Review Efforts l
a.
Maintenance Rule Reauirements:
Administrative controls are established for performance monitoring, trending, and reporting in accordance with 10 CFR 50.65.52 I
The program controls the initiation, analysis, retrieval, trending, and reporting of data related to " Plant Level" and I
" Function Specific" indicators of system performance in j
accordance with 10 CFR 50.65.
By monitoring system performance and taking corrective actions, as necessary, TVA i
ensures that SSC safety-related functions will perform consistent with their design basis requirements.
i b.
System Realth Reports:
The SQN System Health Report is a management reporting tool that is coordinated by each system engineer for each system.
This is a quarterly process that systematically evaluates system performance against established performance goals (e.g., unavailability and reliability).
The evaluation provides a description of known problems and various quantitative performance indicators, including work order backlog, to determine whether or not performance goals have been met or if additional resources are necessary to improve performance.
c.
Operating Experience Programs As previously discussed, TVA's operating Experience Program evaluates experience reports received from NRC, INPO, nuclear vendors and equipment,yuppliers, architect / engineers and constructors.
TVA's review of these reports often results in evaluations that utilize industry experiences to ensure systems, structures, and components are consistent with design basis requirements and will perform as expected.
For example:-
TVA evaluated a concern associated with out of tolerance undervoltage protection relay settings due to test-equipment harmonics as identified by NRC Information Notice 95-05.
To ensure that undervoltage instrumentation performance is being properly verified TVA evaluated the harmonic distortion associated with test equipment used at SQN.
This review determined that some power sources allowed by test instructions exceeded the recommended limits.
TVA identified test equipment that performs within the limits recommended Technical Instruction, 0-TI-SXX-000-004.0, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
53 Nuclear Power Standard 4.4, Managing the operating Experience Program.
42
I in the notice and revised appropriate procedures to require their use when testing these protective relays.
]
TVA evaluated Westinghouse Electric Corporation Infogram 96-02 regarding cross-tied safety injection r
accumulators and the unanalyzed condition that could result.
The SQN design was evaluated to see if a similar concern was applicable.
This evaluation determined that cross-tied accumulators could result in unanalyzed accumulator response at SQN and that this~
configuration should not be utilized.- TVA concluded that the potential to be in this configuration was unlikely because operating procedures did not provide instructions for this type of configuration.
TVA is revising operating procedures as an added measure to provide a caution against cross connecting the accumulators.
d.
Generic Reaulatory Issues TVA also utilizes other line organization reviews to ensure that design basis performance requirements are met.
These reviews, often prompted by NRC or industry initiatives, confirm on a system specific basis that systems, structures, and components are configured consistent with the plant design basis. Examples are provided below.
TVA performed a review of plant design and procedures to address the Station Blackout'(SBO) Rule, (10 CFR 50.63).
This review resulted in the installation of fail-open level control valves for the turbine driven auxiliary feedwater pump (with sufficient air supply for 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> remote operation during a SBO), institution of a site quality assurance program for SBo _ components and revised procedures for t
the emergency diesel generator reliability program.
i TVA performed a review of the SQN post-accident e
zonitoring equipment pursuant to NRC Regulatory Guide (RG) 1.97.
This review identified a number of instrumentation hardware changes requi' red to bring SQN into compliance with the requirements of the RG.
These instrumentation modifications were made by TVA in 1990.
The review also identified approximately 28 deviations from RG 1.97 Standards.
Justifications for these deviations were prepared as part of the review.
These justifications were reviewed by NRC and were-found to be acceptable.
~TVA performed a review of plant design relative to the e
anticipated transients without SCRAM (ATWS) concern (10 CFR 50.62).
This review resulted in the addition of the ATWS mitigation system actuation circuitry 43
(AMSAC) to the plant control system to provide an alternate means of tripping the main turbine and actuating auxiliary feedwater flow independent of the reactor protection system.
The system is designed to actuate when low-low levels are detected in three out of four steam generators coincident with plant power above approximately 40 percent.
This added design feature will prevent reactor coolant system overpressurization, maintain fuel integrity and limit accident radiation releases in the event of a common mode reactor protection system failure.
5.
Independent Assessments:
TVA and NRC assessments of SSCs at SQN provide an additional barometer of whether SSC configuration and performance are consistent with design basis requirements.
a.
TVA Assessments:
In general, the SQN TVA vertical slice reviews have been conducted with emphasis placed on Design and Design Basis, Maintenance, Modifications, and Operations.
These assessments / slices included design document reviews as well as reviews of systems which focused on the analytical basis, design change control, design basis documents, licensing / design interface, configuration and documentation control.
As part of these vertical slices, TVA compared system design' with actual
_ plant configuration and assessed the adequacy of processes that control design.
Through these vertical slices, areas have been identified which could be improved to ensure the highest level of confidence in the system.
Corrective actions have been initiated to address these areas.
Three of these vertical slice assessments are discussed below.
TVA completed a vertical slice inspection of the SQN Main Feedwater, Auxiliary Feedwater, portions of the Safety Injection (accumulators) and attendant electrical power systems in 1996. The TVA inspection team concluded that the plant configuration and Engineering, Modification, Maintenance, and operation programs for the systems reviewed meet regulatory and TVA Nuclear requirements.
The inspection revealed some deficiencies and inconsistencies in design and licensing documents that indicated the implementation of the design control process was not fully effective.
None of the deficiencies identified adversely affected the function of the systems evaluated.' The technical adequacy of the existing plant configurations to perform the required functions were not impacted by the deficiencies.
The deficiencies included certain FSAR discrepancies and minor deficiencies in updating design documents.
TVA is implementing corrective actions that will further enhance the. design control process to help minimize i
future implementation errors.
l i
44 I
I' i
TVA previously performed two vertical slice inspections after the DVBP effort.
In early 1989, a vertical slice was conducted to assess the quality of modifications implemented i
during the SQN Unit 2 Cycle 3 outage and in late 1989, another vertical slice was completed to evaluate the adequacy of the Phase II DBVP.
As part of these vertical
- slices, TVA compared system design with actual plant 4
configuration and assessed the adequacy of processes that
]
control design.
1RTA concluded that overall the DBVP effort was technically l
adequate and complete design criteria were developed, j
modification packages were complete with appropriate design output, controls for the updating of procedures affected by design changes were adequate, and interfaces between the 2
4 design changes and other programs were adequately addressed.
i l
TVA did identify and correct minor weaknesses in plant i
.conformance with design, unverified assumptions not verified j
prior to returning systems to operation, and not updating l
FSAR/ design criteria.
1 i
b.
NRC Assessments:
The previously described NRC Service Water System Operational Performance l
Inspection (SWSOPI) in 1995, identified a violation i
associated with inadequate design control measures for ERCW flood mode, high pressure fire protection (HPFP), and j
emergency diesel _ generator batteries.
Although this violation affected the adequacy of design documents, NRC l
stated that there was no operability impact.
TVA corrected the necessary calculations, revised design output documentation, and updated required drawings in response to the violation.
The NRC inspection included review of design input and output documentation, review of modifications, observation of operator performance by simulator, walkthroughs and interviews, plant walkdowns of configurations, review for adequacy of testing to verify operability, and maintenance histories for-performance of components.
NRC concluded that the essential raw cooling water system was operable and TVA's heat exchanger performance monitoring program was generally acceptable notwithstanding the issues described above.
6.
Summary:
The programs, process, and reviews discussed above provide TVA with assurance that systems,
. structures, and components are configured and will perform in accordance with design.
i 45
D. Reauest (d)
Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence, and reporting to NRC.
i l
TVA ResDOnse to Roguest (d)
There are several ways in which problems are identified.
These include observation by trained personnel, through j
equipment performance, through SQN assessment and audit activities, and through " generic" industry information.
Once identified, problems are placed in TVA's corrective action program for evaluation and correction.
This process is described below along with TVA's reporting processes. The programs at all sites contain, as a minimum, the essential elements described in this response, but there are minor implementation differences between the sites to address specific issues.
It is important to note tbst training received by personnel involved in configuration management, coupled with their j
experience, enhances their ability to identify problems.
Engineering personnel who independently prepare or technically review safety-related and quality-related design changes are initially trained and receive periodic refresher courses.
The operations personnel responsible for configuration control include both NRC-licensed and nonlicensed operators who receive extensive training.
Maintenance and modifications are required to be performed by individuals trained and qualified for each specific task, or under the supervision of a cognizant engineer.
The administratively controlled configuration management process coupled with this training assists the involved personnel in identifying potentially adverse conditions.
In addition to TVA's corrective action program, TVA established an Employee Concerns Program (ECP) to provide an alternative problem reporting mechanism.
The ECP was established to receive, investigate, and respond to concerns raised after February 1, 1986.
The ECP has offices at each operational nuclear plant site and the TVA Nuclear corporate office.
The program name was changed to concerns Resolution (CR) program in 1991.
The program continues to encourage the prompt and effective resolution of concerns through the normal line processes while providing an alternate avenue for concerns that cannot be effectively resolved otherwise.
46
I i
l
}
l 4
1.
TVA Corrective Action Proarant TVA's Corrective
. Action Program contains the processes for the documentation of potential problems, the determination, tracking and j
implementation of corrective actions, including actions to i
determine the extent of problems and to prevent recurrence.
j l
l The SQN Corrective Action Program consists of different
~
processes which document.and correct problems and adverse conditions.
These processes.are designed to address l
problems and adverse conditions in a manner consistent with the nature of the condition and its importance to plant safety..The corrective action program processes applicable to design basis issues are the work request / work i
order (WR/WO) and the Problem Evaluation Report (PER).
The WR/WO is used to identify and correct routine hardware i
i problems or failures on equipment, structures, spare components (i.e., replace packing, correct seat leakage, i
replace motor bearings, etc.).
PERs are used to identify i
and correct nonhardware deficiencies and are also used to address the causec for nonroutine hardware deficiencies.
For example, an unusual component failure that caused damage
_ would result in initiation of both a WR/WO (to fix the specific hardware problem) and a PER (to identify and correct the cause of the hardware problem).
i a.
WR/WO Process:
Routine equipment deficiencies j
are correctgp using the WR/WO process which, in general, is as follows:
A WR card is completed by the initiator to describe the j
deficiency, the equipment involved, and the location of i
the equipment.
The initiator's supervisor reviews the WR card to determine if a PER is also needed.
If so, a PER is initiated.
The initiator or supervisor then submits the WR card to the Operations shift Manager (or his designee) for operability and reportability evaluations.
The Operations Shift Manager (or designee) performs an operability evaluation as soon as possible.
The evaluation must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of WR initiation.
This evaluation ensures that Technical Specification LCOs are reviewed and applied as appropriat%,
The Shift Manager also determines if the deficiency is reportable to NRC or other agencies, and N
TVA allows certain minor maintenance activities that are commensurate with craft qualifications and require little coordination j
to be performed under less restrictive controls than those described in this section.
i 47
i i
initiates any required actions (TVA's reporting process is described further below.).
Finally, the Shift i
e i
Manager assigns a priority to the WR.
i l
The Work Control Group enters the WR into the tracking 4
system and forwards it to the Planning Group.
I f
The Planning Group plans the work, producing a Wo, j
which is subsequently used by maintenance / modification craft to correct the equipment deficiency.
1 i
Work performed is documented on the Wo, and the Wo is closed.
Appropriate data from the Wo is entered into an equipment history database.
l
}:
The equipment history database is periodically i
evaluated to identify recurring equipment problems and other negative trends.
PERs are generated as j
appropriate, and results.are used in the site self-l assessment process described below.
(
i b.
PER Process:
The PER is an important part of the SQN Corrective Action Program because it is the method j
by which root causes, extent of condition, and recurrence control are determined for significant problems.
The PER process is managed by the Nuclear Assurance and t
Licensing (NAEL).
NA&L maintains a database that tracks 1
each individual PER, the development of corrective actions, I
'the schedule and completion of the corrective actions, and j
t the closure of the PER.
Time limits are maintained for i
initiation and review of PERs, as well as development of j
Corrective Action Plans and verification of closures.
NAEL i
monitors the completion of these activities and if the time limits are not met (or appropriately extended) for Level A and B PERs (PER levels are described below), NA&L escalates the matter to management.
NAEL may escalate late actions on Level C PERs.
In general terms, the PER process includes the following:
A PER can be initiated by an employee for any condition, and immediate action is taken as necessary.
Immediate actions may be necessary to protect plant personnel or plant equipment, or if the condition potentially affects operability.
After initiation, the PER undergoes supervisory and/or management review to ensure that any necessary immediate actions are taken, and to assign an organization to investigate the PER.
In addition, the supervisory / management reviewers assign one of four 48 a
. ~ _. -. - -
r i
levels of significance to the PER, based on the following definitions:
=> Level A - Significant Adverse Conditions.
These include:
l l
A major safety-related or QA program condition that has l
occurred with a frequency as to indicate that past recurrence control has been lacking or ineffective.
I Confirmed' adverse trends in quality activities identified by trend analysis.
A programmatic breakdown which negates quality controls or places doubt on the integrity of the affected program.
i Repetitive or deliberate occurrences of procedur-0.
violations that have a direct and detrimental vffect on safety or quality.
Conditions which impact the plant's ability to mitigate I
design basis accidents.
i
=> Level B - Adverse Coaditions that do not meet the Level A significance criteria but are not routine.
These include:
i Quality-related deficiencies which require identification of apparent cause and action to correct the condition in accordance with the Nuclear Quality Assurance Plan.
Human error (inappropriate actions) which could have, under different circumstances, caused a significant plant event or personnel injury.
Responses to regulatory identified issues which did not result from a Level A event.
l l
Recurring events not classified as significant which retain the potential for causing a plant event or personnel injury.
Events or conditions which require root cause analysis -
to support required recurrence control.
These include Licensee Event Reports, NRC violations, and audit findings.
=> Level C - Routine issues.
These include:
49
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I
- Conditions which do not meet the criteria in Levels A or B-but do identify a problem which warrants tracking i
to closure.
Conditions which do not affect operability and are not reportable.
l j
=> Level D - Minor issues.
These include:
Conditions which do not meet the criteria in Levels A, l
3, or C and immediate actions taken were sufficient to l
resolve the condition.
i Conditions which do not affect operability, are not l
reportable, and are not potentially generic.
4 i
PER conditions that are determined by the initiator or e
l the supervisory / management reviewers to potentially j
affect operability or be potentially reportable to NRC are promptly identified to the Operations Shift Manager i
for evaluation.
The Shift Manager determines 1
operability of the affected system or component based on a review of Technical Specification requirements.
This includes evaluation of necessary attendent equipment such as instrumentation, controls, and power supplies.
The ultimate decision on operability rests with the duty Shift Manager.
However, in order to make this operability determination, the Shift Manager may call upon the various engineering resources available on site.
TVA has administrative controls for the performance of engineering gyaluations in support of operability determinations.
These Technical Operability Evaluations (TOES) are performed by Engineering at the request of the Shift Manager in order to obtain formal engineering input for aid in determining operability.
TOES may be initiated to
]
evaluate a past operability concern for reportability purposes, a future operability concern in anticipation j
of an upcoming plant evolution, or for other reasons as requested by the Shift Manager.
The reportability process is discussed later in this response to Request (d).
The organization assigned to investigate the PER nondition (the Responsible Organization) determines the ccuse(s) of the condition and formulates corrective 4
acticna.
Root'cause analyses are required for Level A PER1, when requested by the management reviewers, and for PERs written to address Licensee Event Reports, Notices of Violations, or Quality Assurance Audit i
55 site standerd Practice, SSP 3.4, Corrective Action.
50
.____m.__.
Findings.
The root cause analysis method may be specified by the management reviewer (s).
Otherwi'se, it is selected by the Responsible Organization based on the nature of the condition.
TVA has guidelines for performing yarious types of root cause analysis, including:3
=> Task Analysis
=> Change Analysis l
=> Barrier Analysis
=> Event and causal Factor Charting
=> Advanced Analytical Methods (e.g., Kepner-Tregoe problem solving)
Training on root cause techniques is provided to i
personnel who perform root cause analyses.
i l
Significant events are investigated by multidisciplinary teams to facilitate comprehensive, accurate, and timely root cause analysis.. Once the root cause(s) is determined, the Responsible organization defines corrective actions to remove the cause and thereby prevent recurrence of the condition.
Corrective actions for PERs for which root causes are not required to be determined are based on an apparent cause determination where the specific problem is corrected and data is collected and used for trend analyses.
For example, a single example of a procedure error that has negligible consequences may not warrant a root cause analysis; the corrective action would be to correct the procedure.
However, a series of procedure errors over a period of time discovered through trend analysis may indicate a more significant problem for which a root cause analysis should be done.
Corrective actions, whether from root or apparent cause analyses, are assigned to organizations and tracked to completion.
During the investigation of Level A and B PER conditions, the Responsible Organization determines the extent of the condition.
This determination uses results of the root or apparent cause analysis to identify if other plant programs, processes, or hardware are subject to the same PER condition.
For example, if the PER documents a low flow condition found to be caused by a manufacturing defect in a pump 56 TVA Nuclear Business Practice - 236, " Event Critique and Root Cause Analysis.
51 l
impeller, the extent of condition process would.
determine where else in the plant that impeller type is used.
Appropriate actions would then be taken to address the potentially defective parts.
If a Level A or B condition is determined to be potentially applicable to TVA sites other than the site where it was generated, the PER will be transmitted to the other sites for review.
If those reviews conclude that the condition also exists at the other sites, a new PER is generated at each affected site and is cross-referenced to the initiating PER.
If review by the other sites concludes-that the condition does not
, exist there, a justification for the conclusion is documented.
TVA requires that.significant adverse conditions be processed as level A PERs.
The NA&L organization approves corrective action plans and verifies that corrective actions for level A PERs have been completed as described in the corrective action plan.
This independent verification occurs after the Responsible Organization reports that all actions are complete and before the PER is closed.
The Responsible Organization is required to resolve any problems identified during this verification.
A subsequent effectiveness review is also performed for level A PERs.
After corrective actions have been in place long enough to have removed the cause(s) of the PER condition, the Responsible Organization assesses whether the original corrective actions were effective.
If the corrective action was not effective, a new PER will be generated, c.
self-Assessments:
Self-assessments are performed to identify undesirable changes in personnel, equipment, program, and process performance over time.
Self-assessments center around the development of the quarterly Level I Trend Analysis Report.
The process for generation of this report i.wesives extensive line organizational input.
The report format is patterned after the' management areas in the Institute for Nuclear Power Operations (INPO)90-015 " Performance Objectives and Criteria for Operating and Near-Term Operating License Plants." Each management area is assigned a " window."
Performance for the quarter is assessed by the responsible line organization.
Information such as PERs, Notices of Violation, Licensee Event Reports, performance indicators, recurring equipment problems, and other pertinent data are used to determine the overall performance for each management area.
The corresponding window is then assigned 52
i f
i a color which indicates the performance in that area.
The colors and corresponding performance ratings are as follows:
l Color Performance Ratina Interoretation i
Red Significant Weakness Requires immediate management attention.
j
. Yellow Improvement Needed Requires additional l
l management attention.
4 l
White Satisfactory Meets current standards.
)
i Performance Green Significant Strength Performance exceeding standards or expectations i
i The colors assigned to the windows are then reviewed through i
j a series of site trend analysis committee (STAC) meetings, i
Performance assessment is challenged by peer managers in these. meetings, ensuring that each organization's management i
is'self-critical and is assessing performance to the correct i
standards.
Once complete, the total report is reviewed by a i
)
Trend Review Board (TRB), chaired by the Site Vice President
}
or Plant Manager.
Each site's report is then transmitted to j
corporate headquarters where it is compiled and reviewed by i
TVA senior management.in the Management Trend Oversight i
Board (MTOB).
At each stage, management reviews and challenges the performance ratings to ensure that proper performance standards are applied.
When advarse trends are identified by this process PERs are generated to address the concern.
For example, a trend was identified associated with an excessive number of PERS regarding nonengineered changes to the plant at SQN.
PERs were generated to address this adverse trend and provide common cause evaluations and appropriate corrective actions.
d.
Operatina Ernerience Proaram:
TVA's Operating Experience (OE) program assures that operating information pertinent to plant safety is revigyed and distributed in a timely manner to plant personnel.
Information reviewed by the OE program includes NRC information notices, INPO Significant Operating Experience Reports, INPO Significant
. Event Reports, 10 CFR 21 reports that originate outside TVA, General Electric Services Information Letters, Westinghouse _
Technical Bulletins, and TVA's NRC violation notices.
The applicability of the item is assessed and organizations that could be affected by the experience information are identified.
As applicable, reports are distributed for N
Nuclear Power Standard 4.4, Managing the operating Experience Program.
53
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i h
j information or assigned as action items for evaluation to j
the appropriate TVA plants and organizations.
If these organizations determine that an adverse condition exists, then a PER.is written and the problem is resolved within the corrective action program.
If no adverse condition exists, l
the OE information may result in enhancements to programs,-
l processes, hardware. etc., in order to avoid future j
problems.
Due dates for evaluation of OE documents are established commensurate with the probability and potential
[
impact to the plant.
The action items are tracked until i
j completion i
In addition to the systematic review of industry cperating
}
experience described above, TVA participates in.various industry groups (e.g., Westinghouse Electric Corporation i
Owner's Group) where common problems and initiatives are j-discussed and evaluated.
These groups provide another j
mechanism for communication of industry operating experience.
I e.
Event Reportina Process:
Conditions determined to be potentially reportable are processed in accordance with an administrative procedure that details tgp l
specific reporting requirements in 10 CFR 50.72 and 50.73.
During the initial stages of an event, Operations either identifies or is notified o'f the potential problem. When time is available, the assessment of a problem against the reporting criteria routinely involves other organizations such as Licensing, Site Engineering,_and other organizations
]
that have responsibility for the system, structure, component, or process affected by the potential problem.
Through this method, personnel technically experienced with this type of plant problem provide input into the reportability decision.
The final reporting decision rests with the operations Shift Manager.
The site Licensing organization writes Licensee Event Reports (LERs)'as required by 10 CFR 50.7gg using guidelines contained in an administrative procedure.
The Licensing organization obtains information required for the LER from the corrective action program.
For example, the event description, root cause, and corrective actions to prevent recurrence are generated by the PER process.
The site Licensing organizatior also manages the reporting of defects in basic components and failures to comply with NRC requirements in accordance with 10 CFR 21.
Many of the events or conditions which occur are not clearly discernible as to whether a reportable condition exists.
A 58 Nuclear Power Standard 4.5, Regulatory Reporting Requirements.
59 Ibid.
54 i
Where these cases exist, conservative decision making is exercised.
Recent examples of this conservative reporting practice are identified in the notifications made for an invalid Phase B isolation signal generated while the unit was in cold shutdown and the notification of an engineered safety' feature signal when a main feedwater pump trip signal was generated prior to pump operation.
These reportable events and conditions are typically recorded under the corrective action program.
f.
Informal Reportings In addition to the formal reporting mechanisms, the NRC is apprised of developing plant issues through other communication channels.
For example, the NRC Resident Inspector attends plant morning meetings where developing issues are discussed.
Region II and NRR personnel are advised of reported conditions and plant status, as appropriate.
Routine and event specific communications with the NRC are necessary.to minimize confusion and prevent misunderstandings.
Several vehicles are utilized to ensure appropriate NRC personnel are advised of ongoing efforts and
. findings.
Relative to the resident inspectors, routine communications are held to resolve inspector issues.
Quality Assurance personnel also routinely brief the inspectors on ongoing and planned audits and assessments.
In addition, weekly meetings are held between site management and the resident inspectors to allow for an open forum for direct communication of inspector issues.
Relative to communication with NRC Region II and Nuclear Reactor Regulation management, bi-monthly meetings are currently being held to inform NRC of the status of progress in plant and personnel. performance.
To further facilitate NRC communication, a. Project Plan is issued on a monthly
' basis to communicate TVA's understanding of upcoming inspections, and to identify major submittals and projects which could impact NRC resources.
l i
55
i E.
Reauest (e)
The overall effectiveness of your current processes and programs in concluding that the configuration of your plant (s) is consistent with the design basis.
TVA Response to NRC Request (e)
TVA is confident that its processes and programs have been effective, overall, in ensuring that plant operation, configuration, and performance are consistent with design basis requirements.
The bacia for TVA's confidence has been discussed in the introductory sections of this response and in the specific responses to Requests (a) through (d).
TVA l
is also confident that its program for identifying, l
evaluating, tracking, and resolving problems and adverse conditions is e_ sound program, but recognizes that it has l
not always implemented that program effectively at SQN.
TVA has taken and is continuing to take steps to improve performance in this area at SQN.
TVA recognizes the importance of maintaining plant configuration consistent with design basis requirements and the need to control changes to the design basis to ensure that design basis assumptions remain valid.
TVA also recognizes the importance of maintaining an accurate UFSAR.
TVA continues to be self-critical in these areas and has identified the need for improvements at SQN.
Corrective actions are being implemented to address these improvements.
Discussed below is a summary of the basis for TVA's confidence in its. programs and processes, and a sampling of various " data points" that have measured their effectiveness.
Also discussed below are areas related to design and licensing basis control that TVA has identified l
as needing improvement.
1.
s - =ry of TVA confidence In Desian Basis:
TVA's confidence that the plant configuration and operation is consistent with the design basis and that specific deficiencies are identified and corrected is based upon:
The DBVP, a comprehensive effort that established the i
SQN design basis requirements and reconciled the as-designed and as-constructed facility.
The DBVP also 4
I identified the design basis requirements that were verified by testing; The Restart Test Program, which ensured that l
performance requirements for systems, structures, and components identified by the DBVP were satisfied; 56
i l
i The procedures upgrade program, which ensured that I
design basis requirements were translated into plant procedures; i
r TVA's configuration control processes, designed to ensure changes to the plant configuration are l
thoroughly evaluated and reflected in the design basis i
documentation, the FSAR, and the implementing i
operations, test, and maintenance procedures; Use of the corrective action program by TVA's trained i
and qualified management and line organization
}
personnel to identify and correct problems, including problems related to the design basis; The routine plant and design reviews, and independent and vertical slice assessments, that assess SQN configuration, SSC performance to determine whether they are consistent with the design basis; when problems are identified, corrective actions are taken; and The use of generic induistry and NRC information to review plant design and configuration, as well as plant programs.
As discussed in responses to the information Requests (a) through (d), TVA has already implemented several large-scale efforts to address issues related to design and configuration control for problems identified in the mid 1980s.
TVA has also implemented or initiated a number of other programs, reviews and assessments since that time to ensere that design basis requirements continue to be implemented in procedures and SSC configuration and performance.
2.
Performance /Innlementation Issuest Over the past few years, TVA's assessments of design-related activities have identified the following performance issues for which TVA has taken, or is in the process of taking, corrective actions.
These issues are: plant activities that affected plant design, but that did not undergo the evaluation and review dictated by TVA's design control process; full close out of older design changes; timeliness of corrective actions; and proper dispositioning of abandoned equipment.
These performance issues have not resulted in the identification of conditions that would render safety-related systems inoperable nor have they undermined TVA's overall confidence in its design control process.
But TVA clearly recognizes the need to continue to aggressively pursue and resolve these issues at SQN.
For example, TVA is 57
i i
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i i
f i
currently evaluating and resolving these issues in the areas i
of civil piping and pipe support, and fire protection.
TVA's assessments have also identified the need to improve some licensing-basis controls.
The specific issues i
identified by TVA in this area involve the applicability of 1
10 CFR 50.59, and the overall accuracy of.the UFSAR.
TVA i
identified weaknesses in its 10 CFR 50.59 program that allowed changes to be made to certain nonquality and nonsafety-related procedures and certain design input and outputs without a SA/SE.
No unreviewed safety questions have been identified as a result of these weaknesses.
This program is currently being revised to correct these weaknesses.
TVA is also taking corrective actions to address UFSAR accuracy including verifying the accuracy of the UFSAR as described further below.
3.
Measurements of Effectiveness:
TVA assesses the effectiveness of its control of design basis requirements and the translation of those requirements into procedures and SSC configuration and performance through many specific SSC reviews, and inspections.
Many of these assessments have already been discussed in TVA's responses to Requests (a) through (d).
In addition, to these assessments, TVA, the industry (INPO), and NRC conduct reviews and inspections that also measure the effectiveness of TVA's processes and programs. While TVA does not rely on NRC inspections to determine if plant configuration and operation is consistent with design basis requirements, these inspections do provide and independent barometer of the effectiveness of TVA activities in this area. ' Discussed below are several inspcations and/or reviews conducted by TVA, the industry, and the NRC that have measured the j
effectiveness of TVA's programs.
1 a.
Desian control Process:
As part of the DBVP, TVA assessed its design control process and implemented several changes and enhancements.
The changes and enhancements resulted in an improved design change process to support' restart of the SQN units.
The principal changes made after the DBVP included implementation of a single "as-constructed" drawing program, establishment of single organization design ownership, development of an integrated design change package, and established of a plant management change control board.
The design control process has continued to be enhanced since the DBVP effort to ensure the effectiveness of the process.
In addition, INPO conducted a review of the SQN design control process implementation in 1991.
Through that review, a deficiency was identified that involved instances where the lack of communication resulted in errors associated with the implementation of plant modifications.
)
TVA implemented changes to the design control process to 58
l i
address this issue by including multiple plant organizations in the initial design review for modifications and added an impact review after final package development by site organizations to eliminate miscommunication of design impact.
)
In a separate INPO review conducted in 1995, a concern was identified with the methods for assessing trends in design I
control problems for specific events.
This concern was an irplementation issue rather than a programmatic problem.
specifically, the noted events were the result of process implementation errors and were addressed by providing clarifications to plant personnel regarding the process requirements.
TVA also increased the scope of design-related problems to be considered in self-assessments to ensure that design control issues are properly trended in the self-assessment program at SQN.
The NRC also reviewed TVA's design control process as part j
of SQN's restart after the DBVP was performed.
In NUREG 1232, Volume 2 and Supplement 1, NRC provided the results of their review.
The overall conclusion was that TVA had taken the appropriate steps to correct design control problems at Sequoyah for restart.
NRC stated that TVA's action plan for these improvements represented a significant enhancement to the design control process.
The improvements included such items as review of change requests by a change control board, implementation of standalone modification packages,
-and a drawing system that improves the legibility of controlled drawings along with the use of a single drawing system for primary drawings used to operate the plant.
Since that time, the SQN design control processes have been evaluated by NRC on many occasions.
For example, the NRC's Integrated Performance Assessment Process review of SQN in 1995, as documented in Inspection Report 95-25, found that enhancements in the design control process had adequately resolved identified design contggi issues that had occurred prior to the assessment-period.
In other NRC assessments, the NRC has identified specific cases of failures to properly implement the design control process since the DBVP
]
effort.
However, these are considered individual 2
implementation issues, and are not indicative of any overall programmatic issue.
b.
Corrective Action Procram:
In 1996, TVA conducted an audit of the adequacy and effectiveness of the corrective action program.
The audit team concluded that 4
the' program met the TVA and regulatory requirements such that it is adequate to identify, evaluate, and resolve adverse conditions, however, improved implementation'is NRC Letter to TVA, dated December 21, 1995, Final IPAP Report for SQN, NRC Inspection Report Number 50-327/328 95-25.
59
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required for the program to become fully effective.
TVA i
identified three findings at SQN that indicated a need for improved implementation to ensure that the program is fully effective.
The findings included problem evaluation report l
weaknesses, ineffective recurrence controls, lack of extent of condition analysis, and not evaluating previous or similar occurrences; a weakness in the trend!!.g program to identify ineffective corrective action and recurrence controls; and maintenance history trending of compliance instrumentation deficiencies.
TVA is aggressively pursuing corrective action program implementation issues as discussed above.
Part of TVA's efforts in this area include a t
rigorous self assessment of the corrective action program effectiveness, increased management involvement and overview of the process, an increase in the number of problem evaluation reports that receive a review for previous similar events and extent of condition, improvements in the j
ownership of the program, and stronger quality assurance oversight of the program.
i NRC has also conducted several inspections that have assessed the adequacy of the corrective action program for SQN over the past few years.
In 1995, the NRC conducted an Integrated Performance Assessment Process review of SQN as documented in Inspection Report 95-25.
The adequacy of the
~
self assessment and corrective action activities was evaluated.
The NRC concluded that the areas of problem identification and problem analysis and evaluation were good.
NRC concluded that the problem identification activities continue to be good and improvements have been and continue to be made in analysis activities.
In 1996, the NRC again reviewed the corrective action program at SQN.
In NRC Inspection Report 96-13, the NRC identified a failure to identify the root cause and take adequate corrective actions for recurring failures, a
failure to implement corrective actions to control the use of inappropriate material in safety-related applications, and a failure of extent of conditions reviews to bound affected equipment and provide adequate corrective actions.
These findings are similar to those described above in the previously performed TVA audit associated with the adequacy of the corrective action program.
Again, these issues are focused on the implementation of the corrective action program.
The previously described.TVA and NRC findings are not considared programmatic problems, but instead are implementation problems.
To address these problems, TVA has formed a multidiscipline team to evaluate how the corrective action program is implemented, develop corrective actions, i
if necessary, develop required training, provide coaching for root cause investigations, and transition ownership of the corrective action program to the line organizations.
l 60 1
l
)
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These activities will strengthen the effectiveness of how the corrective action program is used.
c.
Desian Baseline verification Proarant TVA audited and assessed the adequacy of the DBVP.
These evaluations provided TVA with added assurance that the plant configuration and operation are consistent with the design 4
i basis as a result of the DBVP efforts.
l j
In addition, the NRC reviewed the DBVP.
In Sections 2.1, 2.2, and 2.3 of Volume 2 of NUREG-1232, the NRC staff
)
documented its evaluation of the changes to design control, i
the DBVP, and the design calculation program for SQN Unit 2, as described in the Sequoyah Nuclear Performance Plan and related supporting documents.
In Volume 2 of NUREG-1232, the staff concluded that TVA had adequately identified and corrected the problems associated with design control and hadinstgutedanappropriatedesignbasisandverification program.
NRC also concluded that design control problems are being corrected and when completed the plant will-j conform to its licensing basis.
The staff also concluded i
that the design calculations necessary for plant operation had been adequately defined and implemented for restart and appropriate post-restart activities were planned to further enhance other design basis calculations.
The post-modification testing programs were determined to be acceptable.
In NUREG-1232, Volume 2, Supplement 1, the NRC staff a
nMa ons taken M ne redad documented the pa of SQN Unit 1.6 NRC concluded that TVA had resolved the I
necessary SQN specific issues to allow restart on Unit 1.
In addition to the areas described above for Unit 2 restart, NRC continued its review of TVA programs associated with the i
civil calculations and concluded that the calculations were extensively reviewed and inspected and that the program was l
acceptable.
The NRC also conducted several implementation inspections of the DBVP at SQN.
A multidiscipline team of NRC and consultingfirmpersonnelconductedaninspecyonin1987, as documented by NRC Inspection Report 87-31.
The NRC inspection team did not discover any significant discrepancies in TVA's implementation of the DBVP requirements.
Those discrepancies found by NRC were l
61 NRc letter to TVA, dated May 18, 1988, safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nuclear Performance Plan
- NUREG 1232, Volume 2.
62 Safety Evaluation Report on Tennessee Valley Authority:
Sequoyah Nuclear Performance Plan, Sequoyah Unit 1 Restart - NUREG 1232, Volume 2, supplement 1, January 1989.
0 NRC letter to TVA, dated December 3, 1987, Inspection Report Nos. 50-327/328 87-31.
61
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1 promptly addressed by TVA and corrected.
As a result of the
]
inspection, the NRC team concluded that the DBVP was l-generally conducted in accordance with the program plan.
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d.
Procedure Upgrade Program As part of the i
restart plan for SQN Units 1 and 2, TVA implemented a Surveillance Instruction Review and Revision Program and a procedure enhancement program to support the restart of the 4
I units.
These' programs included the revision of procedures j
important to restart or safe operation that required i
revision or development as a result of completed plant modifications and DBVP system walkdowns.
These revisions ensured the appropriate incorporation of design basis i
information.
NRC reviewed these programs and concluded in Volume 2 and the subsequent Supplement 1 of NUREG-1232 that l
TVA was producing adequate procedures to support the startup i
of each unit.
TVA's program for the improvement of procedures included a long-term process as part of a corporate-wide plan that extended beyond the restart of the l
SQN units.
In addition, the use of impact reviews as part j
of the design control process ensures that affected procedures are evaluated and revised to support plant operation and maintain the design basis.
Reviews by NRC of procedure adequacy to date have not resulted in the discovery of deficiencies that would indicate a weakness in l
TVA's ability to maintain procedures that appropriately j
reflect the design basis.
1 l
e.
site Manaaement:
l SQN several times since the DBVP effort.
In 1992 INPO noted j
a " Good Practice" for comprehensive monitoring and i
assessment by management that contributed to improved levels of plant performance.
In the 1994 and 1995 INPO l
evaluations, findings were identified associated with l
communication of expectations to site personnel, lack of i
accountability for activities, and timely corrective actions.
TVA has responded to these findings by implementing actions to improve communications and accountability and requiring periodic reviews of incomplete corrective actions.
i TVA's Quality Assurance organization recently conducted an j
assessment to determine if known degraded and/or j
nonconforming conditions have been promptly corrected.
While the overall program for identifying and correcting these deficiencies was found to be sound, implementation i
weaknesses were identified.
An apparent reason for these I
weaknesses was ineffective site management oversight.
l Corrective actions are being developed to address these weaknesses as well as the controls that are necessary to j
minimize the occurrence of future problems.
Collectively, the above items indicate a concern in the area of management attention to ensure timely resolution of long i
62
i standing problems that could impact the design basis of the i
plant.
Examples of some of the issues that indicate the 1
extent of chis concern includes civil piping and pipe to NRC),gpnconformances (previously documented in a letter support fire protection equipment adverse conditions, and
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abandoned or inoperable radiation monitors.
Corrective actions have been, or are being, developed to address these areas and ensure the SQN design basis is adequately i
maintained.
l 4.
Updated Final Safety Analysis Report:
In addition I
to TVA's self-critical approach to design basis issues, TVA is also assessing licensing basis control.
In this j
connection, TVA has recently embarked upon UFSAR improvement 1
initiatives:
As part of the Unit 2 cycle 7 refueling outage observations, a selected review of ten chapters of the FSAR was performed by Quality Assurance with the focus being to evaluate refueling outage practices, processes, and configuration.
While no safety-j significant problems were identified, discrepancies were found as a result of procedure and practice l
inadequacies with respect to the FSAR descriptions and were documented in a problem evaluation report.
Examples of these discrepancies are instrumentation J
calibration and leak test frequencies were not consistent with the FSAR, a surveillance instruction did not require the locking of valves as the FSAR stated, and containment airlock lighting and communication provisions were inconsistent with the FSAR descriptions.
TVA corrected the discrepancies by evaluating the individual FSAR sections that were affected against the design basis and revising appropriate areas.
In June 1996, TVA initiated the cyclic review of the UFSAR.
This review is performed by both the lead and support organizations for each section of the UFSAR.
Special emphasis was placed on ensuring that the plant is operated as described in the UFSAR.
This review identified discrepancies.
In October 1996, an additional verification of the UFSAR was conducted.
The UFSAR verification plan included an UFSAR/ Technical Specification (TS) consistency review, an UFSAR/ modifications safety assessment / safety evaluation consistency review, an UFSAR/ method of operation consistency review, an 64 TVA letter to NRC dated January 31, 1994, Sequoyah Nuclear Plant - Generic Letter 91 Resolution of Piping and Pipe Support Nonconfonnances.
63
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UFSAR/FSAR change request consistency review, and a QA i
evaluation of the verification process.
The UFSAR/TS and the UFSAR/ modifications consistency review identified a number of discrepancies.
In addition, the UFSAR/ method of operation consistency review consisted of a complete review of UFSAR sections j
containing system or component descriptions by I
i operations.
This review also identified discrepancies.
I Each of the above discussed discrepancies, not i
incorporated in Amendment 12 of the UFSAR, are being
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dispositioned under the Amendment 13 update scheduled to be submitted to NRC by June 6, 1997.
i In November 1996, a detailed vertical slice of the main and auxiliary feedwater system and phe cold leg accumulator systems was conducted.6' This assessment
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identified UFSAR discrepancies similar to those identified in the QA assessment discussed above.
Corrective actions for these discrepancies have been developed and are being implemented in accordance with 4
j the corrective action program.
In December 1996, TVA initiated a specific Quality Assurance assessment of conformance with the UFSAR.66 l
The purpose of the assessment was to determine the i
effectiveness of the line organization's verification l
of the UFSAR accuracy.
This assessment confirmed the l
results of the UFSAR consistency reviews discussed j
above.
The review included an evaluation of three systems -- the reactor coolant system, diesel generator i
fuel oil system, and containment spray system.
The j
system evaluations involved comparisons of selected operating procedures and associated abandoned equipment to UFSAR descriptions.
4 The conclusion of this assessment was that no safety-significant problems or unreviewed safety questions 3
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were identified.
TVA also concluded that the UFSAR l
verification effort was adequate in determining that UFSAR discrepancies existed, however, they did not
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always identify the extent of the discrepancies.
Engineering is continuing with its UFSAR validation
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efforts.
Design change activities also include the i
validation of each UFSAR section affected by the change.
Identified discrepancies are being evaluated i
and required UFSAR changes are being prepared 65 l
Nuclear Assurance and Licensing (NA&L) - Vertical Slice Inspection Audit, NA-sg-96-15, dated November 20, 1996.
Nuclear Assurance and Licensing (NA&L) - Updated Final Safety 66 Analysis Report - Verification Assessment NA-SQ-96-035, dated j
1 d
December 16, 1996.
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64 i
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.. -. -. -. ~.. - ~. -. -.. -. - - = -.. -
J i
i i
The types of discrepancies identified ranged from overall t
typographical / editorial errors to descriptions of obsolete / abandoned equipment and errors in explanations of 4
system operations.
The resolution of these discrepancies has not resulted in Unreviewed Safety Questions, inoperable systems, or required modifications to plant' equipment, nor have they caused TVA to question the validity of the SQN design basis.
The errors that have been identified are being corrected under the corrective action program.
TVA is continuing to review the UFSAR, evaluate the remaining discrepancies, and prepare the required' changes to the UFSAR.
For these changes to the FSAR, TVA will issue Amendment 13 to SQN's UFSAR.
i 65
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l F.
Additional Efforts:
TVA is continuing efforts to assess j
and verify the accuracy of the required design basis for SQN 4
and programs used to maintain that basis.
These efforts are 4
in addition to those programs and processes that continually a
maintain the required design basis.
The activities planned for the future provide a high level of confidence that the SQN design basis will be appropriately assessed and adequately maintained.
The following is a description of some of these efforts.
l 1.
F81R Verifications TVA is currently performing an FSAR verification effort that will further improve the j
accuracy of the.FSAR.
Engineering is continuing with its J
UFSAR validation effort and corrective actions are being developed and implemented for identified discrepancies.
2.
Confirmatory vertical Slice Review Assessments:
As a further measure to confirm the conformance of plant-configuration with design basis requirements, TVA is currently using the vertical slice technique as part of the 1
formal audit program at each site.
Vertical slice audits are performed at TVA's nuclear plant sites and, if applicable, at the corporate office.
The vertical slices are directed at a particular system and incorporate the periodic audits required by 10 CFR 50, Appendix B.
Other required audits, e.g.,
fire protection, emergency preparedness, security, safeguards, ano fitness for duty, will continue to be performed according to their required periodicity, either as part of or separate from the vertical slice reviews.
Vertical slice audits are comprehensive and will evaluate the engineering design and configuration controls related to the selected system, compare the as-built plant and as-i modified condition of the system, verify system performance',
and assess whether design basis requirements for that system have been translated into associated operating, maintenance, and test procedures.
Audit findings will be' documented, tracked, and corrected in accordance with TVA's corrective i
action program.
- 3. Licensina Basis Quality Assurance will perform licensing basis assessments to determine if changes to the design basis are adequately reflected in the licensing basis, as applicable.
The assessments will be tailored to evaluate the SQN programs in place that identify and control commitments that affect the licensing basis.
Planned review areas include design change notices, changes to the Quality Assurance, Security and Emergency Preparedness programs, outstanding corrective actions, operating procedures, FSAR change requests, relief requests, operator workarounds, 66
-.... ~. -.. ~. - -. _.
o l
operations standing orders, and nonconforming items.
Results of this review will help identify missing or i
l incorrectly applied programmatic elements that can lead to licensing basis differences.
t t
4.
Material condition:
TVA is implementing an assessment plan to place additional emphasis on " material i
condition" of the plant.
This effort includes a review of-open work requests and work orders by a multidiscipline team l
This i
to determine if potential items require escalation.
effort will also review system health reports and master issue list to determine adequacy.
Past potential problems j
will be reviewed to ensure appropriate corrective actions i
and recurrence controls have been implemented and to j
identify hidden adverse trends.
Issues identified during this assessment will be corrected to ensure that significant problems have been addressed.
l 5.
corrective Action Proarant Implementation concerns f
are being addressed by TVA through a corrective action improvement plan that has been initiated at SQN.
This plan i
continues the current practice of utilizing senior management review of Level A and B corrective actions.and root cause evaluations, training of front line personnel for their role in the corrective action process, training of
. event investigators in root cause analys s tec n ques, and i
hi improvements of the apparent cause analysis process.
In addition, a multidiscipline team has been formed to validate l
problems with the corrective action program, develop
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corrective actions for valid problems, develop required training, provide coaching for root cause investigations, identify goals, and transition ownership of the corrective j
action program to the line organizations.
These activities l
will strengthen the corrective action program and provide an l
enhanced site awareness regarding implementation expectations.
1 67
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